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Sample records for mixed-oxide fuel properties

  1. Fabrication of uranium-americium mixed oxide fuels: thermodynamical modeling and materials properties

    International Nuclear Information System (INIS)

    Prieur, D.

    2011-01-01

    Fuel irradiation in pressurized water reactors lead to the formation of fission products and minor actinides (Np, Am, Cm) which can be transmuted in fast neutrons reactors. In this context, the aim of this work was to study the fabrication conditions of the U 1-y Am y O 2+x fuels which exhibit particular thermodynamical properties requiring an accurate monitoring of the oxygen potential during the sintering step. For this reason, a thermodynamical model was developed to assess the optimum sintering conditions for these materials. From these calculations, U 1-y Am y O 2+x (y=0.10; 0.15; 0.20; 0.30) were sintered in two range of atmosphere. In hyper-stoichiometric conditions at low temperature, porous and multiphasic compounds are obtained whereas in reducing conditions at high temperature materials are dense and monophasic. XAFS analyses were performed in order to obtain additional experimental data for the thermodynamical modeling refinement. These characterizations also showed the reduction of Am(+IV) to Am(+III) and the partial oxidation of U(+IV) to U(+V) due to a charge compensation mechanism occurring during the sintering. Finally, taking into account the high - activity of Am, self-irradiation effects were studied for two types of microstructures and two Am contents (10 and 15%). For each composition, a lattice parameter increase was observed without structural change coupled with a macroscopic swelling of the pellet diameter up to 1.2% for the dense compounds and 0.6% for the tailored porosity materials. (author) [fr

  2. Critical experiments with mixed oxide fuel

    International Nuclear Information System (INIS)

    Harris, D.R.

    1997-01-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

  3. Plutonium spot of mixed oxide fuel, 2

    International Nuclear Information System (INIS)

    Suzuki, Yukio; Maruishi, Yoshihiro; Satoh, Masaichi; Aoki, Toshimasa; Muto, Tadashi

    1974-01-01

    In a fast reactor, the specification for the homogeneity of plutonium in plutonium-uranium mixed-oxide fuel is mainly dependent on the nuclear characteristics, whereas in a thermal reactor, on thermal characteristics. This homogeneity is measured by autoradiography as the plutonium spot size of the specimens which are arbitrarily chosen fuel pellets from a lot. Although this is a kind of random sampling, it is difficult to apply this method to conventional digital standards including JIS standards. So a special sampling inspection method was studied. First, it is assumed that the shape of plutonium spots is spherical, the size distribution is logarithmic normal, and the standard deviation is constant. Then, if standard deviation and mean spot size are given, the logarithmic normal distribution is decided unitarily, and further if the total weight of plutonium spots for a lot of pellets is known, the number of the spots (No) which does not conform to the specification can be obtained. Then, the fraction defective is defined as No devided by the number of pellets per lot. As to the lot with such fraction defective, the acceptance coefficient of the lot was obtained through calculation, in which the number of sampling, acceptable diameter limit observed and acceptable conditions were used as parameters. (Tai, I.)

  4. Fabrication of mixed oxide fuel using plutonium from dismantled weapons

    International Nuclear Information System (INIS)

    Blair, H.T.; Chidester, K.; Ramsey, K.B.

    1996-01-01

    A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab

  5. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

  6. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

  7. Introduction of mixed oxide fuel elements in the belgian cores

    International Nuclear Information System (INIS)

    Charlier, A.F.; Hollasky, N.A.

    1994-01-01

    The important amount of plutonium recovered from the reprocessing of spent fuel on the one hand, the national and international experience of the use of mixed oxide UO 2 -PuO 2 fuel in power reactors on the other hand, have led Belgian utilities to decide the introduction of Mixed-Oxide fuel in Doel unit 3 and Tihange unit 2 cores. The 'MOX' project has shown that it was possible without reducing safety or requiring modifications of the plant equipment. It has been approved by the Belgian 'Nuclear Safety Commission'. (authors). 1 tab., 2 figs

  8. Reirradiation of mixed-oxide fuel pins at increased temperatures

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, E.T.

    1976-05-01

    Mixed-oxide fuel pins from EBR-II irradiations were reirradiated in the General Electric Test Reactor (GETR) at higher temperatures than experienced in EBR-II to study effects of the increased operating temperatures on thermal/mechanical and chemical behavior. The response of a mixed-oxide fuel pin to a power increase after having operated at a lower power for a significant portion of its life-time is an area of performance evaluation where little information currently exists. Results show that the cladding diameter changes resulting from the reirradiation are strongly dependent upon both prior burnup level and the magnitude of the temperature increase. Results provide the initial rough outlines of boundaries within which mixed-oxide fuel pins can or cannot tolerate power increases after substantial prior burnup at lower powers

  9. Description of a reference mixed oxide fuel fabrication plant (MOFFP)

    International Nuclear Information System (INIS)

    1978-01-01

    In order to evaluate the environment impact, due to the Mixed Oxide Fuel Fabrication Plants, work has been initiated to describe the general design and operating conditions of a reference Mixed Oxide Fuel Fabrication Plant (MOFFP) for the 1990 time frame. The various reference data and basic assumptions for the reference MOFFP plant have been defined after discussion with experts. The data reported in this document are only made available to allow an evaluation of the environmental impact due to a reference MOFFP plant. These data have therefore not to be used as recommandation, standards, regulatory guides or requirements

  10. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-06-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

  11. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-01-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

  12. Safety criteria related to microheterogeneities in LWR mixed oxide fuels

    International Nuclear Information System (INIS)

    Renard, A.; Mostin, N.

    1978-01-01

    The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

  13. Safety problems related to microheterogeneities in physically mixed oxide fuels

    International Nuclear Information System (INIS)

    Renard, A.; Evrard, G.; Vanhellemont, G.

    1976-01-01

    The safety aspects of microheterogeneities in LMFBR mixed oxide fuel are reviewed from the point of view of the pin behaviour dynamic study, the fabrication and the quality control. The paper emphasizes some significant parameters in transient conditions, the prevention means in the fabrication process and the analysis methods for the control

  14. High temperature transient deformation of mixed oxide fuels

    International Nuclear Information System (INIS)

    Slagle, O.D.

    1986-01-01

    The purpose of this paper is to present recent experimental results on fuel creep under transient conditions at high temperatures. The effect of temperature, stress, heating rate, density and grain size were considered. An empirical formulation is derived for the relationship between strain, stress, temperature and heating rate. This relationship provides a means for incorporating stress relief into the analysis of fuel-cladding interaction during an overpower transient. The effect of sample density and initial grain size is considered by varying the sample parameters. Previously derived steady-state creep relationships for the high temperature creep of mixed oxide fuel were combined with the time dependency of creep found for UO 2 to calculate a transient creep relationship for mixed oxide fuel. These calculated results were found to be in good agreement with the measured high temperature transient creep results

  15. Dissolution of mixed oxide fuel as a function of fabrication variables

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation

  16. Passive measurements of mixed-oxide fuel for nuclear nonproliferation

    International Nuclear Information System (INIS)

    Dolan, Jennifer L.; Flaska, Marek; Pozzi, Sara A.; Chichester, David L.

    2013-01-01

    We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements

  17. Mixed oxide fuel pellet and manufacturing method thereof

    International Nuclear Information System (INIS)

    Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

    1993-01-01

    In a method of manufacturing nuclear fuel pellets which comprises compression molding a mixed oxide powder containing UO 2 and PuO 2 followed by sintering, a sintering agent having a composition comprising about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 is mixed to a mixed oxide at a ratio of about 40ppm to about 0.5 wt% based on the total amount of the mixed oxide and the sintering agent, to prepare a mixture. The mixture is molded into a compression product and then sintered at a weakly acidic atmosphere at a temperature of about 1500degC to 1800degC. With such procedures, the sintering agent forms an eutectic product of a single liquid phase, PuO 2 is dispersed over the entire region of the pellet by way of the liquid phase, formation of a solid solution phase is promoted to annihilate a free PuO 2 phase. Further, growth of crystal grains is promoted. Accordingly, since the MOX fuel pellets prepared according to the present invention have a uniform solid solution state, and no free PuO 2 phase remains, increase of FP gas emission due to local nuclear fission of Pu can be avoided. (T.M.)

  18. Emission computer tomography on a Dodewaard mixed oxide fuel pin

    International Nuclear Information System (INIS)

    Buurveld, H.A.; Dassel, G.

    1993-12-01

    A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

  19. Equipment to weld fuel rods of mixed oxides

    International Nuclear Information System (INIS)

    Aparicio, G.; Orlando, O.S.; Olano, V.R.; Toubes, B.; Munoz, C.A.

    1987-01-01

    Two welding outfits system T1G were designed and constructed to weld fuel rods with mixed oxides pellets (uranium and plutonium). One of them is connected to a glove box where the loading of sheaths takes place. The sheaths are driven to the welder through a removable plug pusher in the welding chamber. This equipment was designed to perform welding tests changing the parameters (gas composition and pressure, welding current, electrode position, etc.). The components of the welder, such as plug holder, chamber closure and peripheral accessories, were designed and constructed taking into account the working pressures in the machine, which is placed in a controlled area and connected to a glove box, where special safety conditions are necessary. The equipment to weld fuel bars is complemented by another machine, located in cold area, of the type presently used in the fuel elements factory. This equipment has been designed to perform some welding operations in sheaths and mixed oxide rods of the type Atucha I and II. Both machines have a programmed power supply of wide range and a vacuum, and pressurizing system that allows the change of parameters. Both systems have special features of handling and operation. (Author)

  20. Transmutation of minor actinide using BWR fueled mixed oxide

    International Nuclear Information System (INIS)

    Susilo, Jati

    2000-01-01

    Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

  1. Some alternatives to the mixed oxide fuel cycle

    International Nuclear Information System (INIS)

    Deonigi, D.E.; Eschbach, E.A.; Goldsmith, S.; Pankaskie, P.J.; Rohrmann, C.A.; Widrig, R.D.

    1977-02-01

    While on initial examination each of the six fuel cycle concepts (tandem cycle, extended burnup, fuel rejuvenation, coprocessing, partial reprocessing, and thorium) described in the report may have some potential for improving safeguards, none of the six appears to have any other major or compelling advantages over the mixed oxide (MOX) fuel cycle. Compared to the MOX cycle, all but coprocessing appear to have major disadvantages, including severe cost penalties. Three of the concepts-tandem, extended burnup, and rejuvenation--share the basic problems of the throwaway cycle (GESMO Alternative 6): without reprocessing, high-level waste volumes and costs are substantially increased, and overall uranium utilization decreases for three reasons. First, the parasitic fission products left in the fuel absorb neutrons in later irradiation steps reducing the overall neutronic efficiencies of these cycles. Second, discarded fuel still has sufficient fissile values to warrant recycle. Third, perhaps most important, the plutonium needed for breeder start-up will not be available; without the breeder, uranium utilization would drop by about a factor of sixty. Two of the concepts--coprocessing and partial reprocessing--involve variations of the basic MOX fuel cycle's chemical reprocessing step to make plutonium diversion potentially more difficult. These concepts could be used with the MOX fuel cycle or in conjunction with the tandem, extended burnup and rejuvenation concepts to eliminate some of the problems with those cycles. But in so doing, the basic impetus for those cycles--elimination of reprocessing for safeguards purposes--no longer exists. Of all the concepts considered, only coprocessing--and particularly the ''master blend'' version--appears to have sufficient promise to warrant a more detailed study. The master blend concept could possibly make plutonium diversion more difficult with minimal impact on the reprocessing and MOX fuel fabrication operations

  2. Dissolution of mixed oxide spent fuel from FBR

    International Nuclear Information System (INIS)

    Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

    1991-01-01

    At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

  3. Modern methods of material accounting for mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

    1981-01-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

  4. Sensitivity analysis of characteristics of spent mixed oxide fuel

    International Nuclear Information System (INIS)

    Hagura, Naoto; Yoshida, Tadashi

    2008-01-01

    Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 σ basis. (authors)

  5. Sensitivity analysis of characteristics of spent mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hagura, Naoto; Yoshida, Tadashi [Musashi Institute of Technology, 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan)

    2008-07-01

    Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 sigma basis. (authors)

  6. Fuel-cladding chemical interaction in mixed-oxide fuels

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, J.W.; Devary, J.L.

    1978-10-01

    The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

  7. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

  8. Development of an engineered safeguards system concept for a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Chapman, L.D.; de Montmollin, J.M.; Deveney, J.E.; Fienning, W.C.; Hickman, J.W.; Watkins, L.D.; Winblad, A.E.

    1976-08-01

    An initial concept of an Engineered Safeguards System for a representative commercial mixed-oxide fuel fabrication facility is presented. Computer simulation techniques for evaluation and further development of the concept are described. An outline of future activity is included

  9. Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, J.W.; Dutt, D.S.

    1976-01-01

    An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

  10. Fission gas behavior in mixed-oxide fuel during transient overpower

    International Nuclear Information System (INIS)

    Randklev, E.H.; Treibs, H.A.; Mastel, B.; Baldwin, D.L.

    1979-01-01

    Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents/s) tested mixed-oxide fuel pins and their steady state siblings

  11. Breaking up of pure and simulated 'burnt' mixed oxide fuel by chemical interaction with oxidized sodium

    International Nuclear Information System (INIS)

    Besnard, R.; Chaudat, J.P.

    1983-01-01

    A large experimental program have permitted to investigate the behaviour of mixed oxide fuel coming in contact with hot oxidized sodium. The kinetic of the reaction, the size and the chemical nature of the particules after interaction have been studied. The main part of experiments have been performed using mixed oxide fuel non irradiated at first and with simulated fission products afterwards. Complementary informations have been obtained with UO 2 fuel pellets. After description of the experimental devices, the results are discussed and the importance of the main parameters, like temperature and fission products effect, are pointed out. (orig.)

  12. Fuel-sodium reaction product formation in breached mixed-oxide fuel

    International Nuclear Information System (INIS)

    Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

    1988-01-01

    The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

  13. Modern methods of material accounting for mixed-oxide fuel-fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Brouns, R.J.; Brite, D.W.; Pindak, J.L.

    1981-07-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MC and A) philosophy have been applied to a mixed-oxide fuel-fabrication plant to produce a detailed preliminary MC and A system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed-oxide fuel-fabrication plant study. Topics covered in this paper include: mixed-oxide fuel-fabrication process description, process disaggregation into MC and A system control units, quantitative results of analysis of control units for abrupt and recurring loss-detection capability, impact of short- and long-term holdup on loss-detection capability, response to alarms for abrupt loss, and response to alarms for recurring loss

  14. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

    1977-04-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  15. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  16. Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

    1977-01-01

    The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

  17. Safety aspects related to burnup increase and mixed oxide fuel

    International Nuclear Information System (INIS)

    Thomas, W.

    1992-01-01

    The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

  18. Behavior of mixed-oxide fuel subjected to multiple thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Neimark, L.A.; Poeppel, R.B.; Hofman, G.L.

    1985-01-01

    The microstructural behavior of irradiated mixed-oxide fuel subjected to multiple, mild thermal transients was investigated using direct electrical heating. The results demonstrate that significant intergranular porosity, accompanied by large-scale (>90%) release of the retained fission gas, developed as a result of the cyclic heating. Microstructural examination of the fuel indicated that thermal-shock-induced cracking of the fuel contributed significantly to the increased swelling and gas release. 29 refs., 12 figs

  19. Behavior of mixed-oxide fuel subjected to multiple thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Hofman, G.L.; Neimark, L.A.; Poeppel, R.B.

    1983-11-01

    The microstructural behavior of irradiated mixed-oxide fuel subjected to multiple, mild thermal transients was investigated using direct electrical heating. The results demonstrate that significant intergranular porosity, accompanied by large-scale (>90%) release of the retained fission gas, developed as a result of the cyclic heating. Microstructural examination of the fuel indicated that thermal-shock-induced cracking of the fuel contributed significantly to the increased swelling and gas release

  20. Radial plutonium redistribution in mixed-oxide fuel

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Schwinkendorf, K.N.; Karnesky, R.A.

    1981-10-01

    Alpha autoradiographs from all HEDL fuel pin metallography samples are evaluated and catalogued according to different plutonium distribution patterns. The data base is analyzed for effects of fabrication and operating parameters on redistribution

  1. In Vivo Calibration Measurements for Mixed Oxide Nuclear Reactor Fuel

    National Research Council Canada - National Science Library

    Shaw, Daniel

    2003-01-01

    .... Thus, the final isotopic composition of MOX depends upon the origin of the plutonium, how well chemical and radiological impurities are removed from the plutonium prior to fuel fabrication, the time...

  2. 76 FR 65544 - Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities

    Science.gov (United States)

    2011-10-21

    ... NUCLEAR REGULATORY COMMISSION [NRC-2009-0323] Standard Format and Content of License Applications... revision to regulatory guide (RG) 3.39, ``Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities.'' This guide endorses the standard format and content for license...

  3. Design of an engineered safeguards system for a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Winblad, A.E.; McKnight, R.P.; Fienning, W.C.; Fenchel, B.R.

    1977-06-01

    Several Engineered Safeguards System concepts and designs are described that provide increased protection against a wide spectrum of adversary threats. An adversary sequence diagram that outlines all possible adversary paths through the safeguards elements in a mixed-oxide fuel fabrication facility is shown. An example of a critical adversary path is given

  4. Used mixed oxide fuel reprocessing at RT-1 plant

    Energy Technology Data Exchange (ETDEWEB)

    Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

    2016-07-01

    Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

  5. Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

    1993-12-01

    A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

  6. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  7. Molten carbonate fuel cell cathode with mixed oxide coating

    Science.gov (United States)

    Hilmi, Abdelkader; Yuh, Chao-Yi

    2013-05-07

    A molten carbonate fuel cell cathode having a cathode body and a coating of a mixed oxygen ion conductor materials. The mixed oxygen ion conductor materials are formed from ceria or doped ceria, such as gadolinium doped ceria or yttrium doped ceria. The coating is deposited on the cathode body using a sol-gel process, which utilizes as precursors organometallic compounds, organic and inorganic salts, hydroxides or alkoxides and which uses as the solvent water, organic solvent or a mixture of same.

  8. Behavior of molybdenum in mixed-oxide fuel

    International Nuclear Information System (INIS)

    Giacchetti, G.; Sari, C.

    1976-01-01

    Metallic molybdenum, Mo--Ru--Rh--Pd alloys, barium, zirconium, and tungsten were added to uranium and uranium--plutonium oxides by coprecipitation and mechanical mixture techniques. This material was treated in a thermal gradient similar to that existing in fuel during irradiation to study the behavior of molybdenum in an oxide matrix as a function of the O/(U + Pu) ratio and some added elements. Result of ceramographic and microprobe analysis shows that when the overall O/(U + Pu) ratio is less than 2, molybdenum and Mo--Ru--Rh--Pd alloy inclusions are present in the uranium--plutonium oxide matrix. If the O/(U + Pu) ratio is greater than 2, molybdenum oxidizes to MoO 2 , which is gaseous at a temperature approximately 1000 0 C. Molybdenum oxide vapor reacts with barium oxide and forms a compound that exists as a liquid phase in the columnar grain region. Molybdenum oxide also reacts with tungsten oxide (tungsten is often present as an impurity in the fuel) and forms a compound that contains approximately 40 wt percent of actinide metals. The apparent solubility of molybdenum in uranium and uranium--plutonium oxides, determined by electron microprobe, was found to be less than 250 ppM both for hypo- and hyperstoichiometric fuels

  9. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

  10. Cesium relocation in mixed-oxide fuel pins resulting from increased temperature reirradiation

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Woodley, R.E.; Weber, E.T.

    1976-06-01

    Mixed-oxide fuel pins from EBR-II test subassemblies PNL-3 and PNL-4 were reirradiated in the GETR to study effects of increased fuel and cladding temperatures on chemical and thermomechanical behavior. Radial and axial distributions of cesium were obtained using postirradiation nondestructive precision gamma-scanning techniques. Data presented relate to the dependence of cesium distribution and transport processes on temperature gradients which were altered after substantial steady-state operation

  11. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

  12. Simulated physical inventory verification exercise at a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Reilly, D.; Augustson, R.

    1985-01-01

    A physical inventory verification (PIV) was simulated at a mixed-oxide fuel fabrication facility. Safeguards inspectors from the International Atomic Energy Agency (IAEA) conducted the PIV exercise to test inspection procedures under ''realistic but relaxed'' conditions. Nondestructive assay instrumentation was used to verify the plutonium content of samples covering the range of material types from input powders to final fuel assemblies. This paper describes the activities included in the exercise and discusses the results obtained. 5 refs., 1 fig., 6 tabs

  13. Dissolution test for homogeneity of mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Experiments were performed to determine the relationship between fuel pellet homogeneity and pellet dissolubility. Although, in general, the amount of pellet residue decreased with increased homogeneity, as measured by the pellet figure of merit, the relationship was not absolute. Thus, all pellets with high figure of merit (excellent homogeneity) do not necessarily dissolve completely and all samples that dissolve completely do not necessarily have excellent homogeneity. It was therefore concluded that pellet dissolubility measurements could not be substituted for figure of merit determinations as a measurement of pellet homogeneity. 8 figures, 3 tables

  14. Decommissioning of a mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Buck, S.; Colquhoun, A.

    1990-01-01

    Decommissioning of the coprecipitation plant, which made plutonium/uranium oxide fuel, is a lead project in the BNFL Sellafield decommissioning programme. The overall programme has the objectives of gaining data and experience in a wide range of decommissioning operations and hence in this specific project to pilot the decommissioning of plant heavily contaminated with plutonium and other actinides. Consequently the operations have been used to test improvements in temporary containment, contamination control and decontamination methods and also to develop in situ plutonium assay, plutonium recovery and size-reduction methods. Finally the project is also yielding data on manpower requirements, personnel radiation uptake and waste arisings to help in the planning of future decommissioning projects

  15. Ion sensing properties of vanadium/tungsten mixed oxides

    International Nuclear Information System (INIS)

    Guidelli, Eder Jose; Guerra, Elidia Maria; Mulato, Marcelo

    2011-01-01

    Vanadium/tungsten mixed oxide (V 2 O 5 /WO 3 ) sensing membranes were deposited on glassy carbon substrates and used as the H + sensor of the extended gate field effect transistor (EGFET) device. X-ray diffractograms indicated a decrease of the interplanar spacing of V 2 O 5 after the insertion of WO 3 revealing that the lamellar structure is under compressive stress. The crystallinity increases with increasing WO 3 molar ratio. The film is not homogeneous, with more WO 3 material sitting at the surface. This influences the response of pH sensors using the EGFET configuration. The maximum sensitivity of 68 mV pH -1 was obtained for the sample with 5% WO 3 molar ratio. For higher WO 3 molar ratios, the behavior is not linear. It can be concluded that V 2 O 5 dominates for acidic solutions while WO 3 dominates for basic solutions. Therefore, the mixed oxide with low amount of WO 3 is the main candidate for further use as biosensor.

  16. Fabrication experience with mixed-oxide LWR fuels at the BELGONUCLEAIRE plant

    International Nuclear Information System (INIS)

    Vanhellemont, G.

    1979-01-01

    For nearly 20 years BELGONUCLEAIRE has been involved in a steadily growing effort to increase its production of mixed oxides. This programme has ranged from basic research and process development through a pilot-scale unit to today's mixed-oxide fuel fabrication plant at Dessel, which has been in operation for just over 5 years. The reference fabrication flow sheet includes UO 2 , PuO 2 and a scraped powder preparation, sintered ground pellets as well as rod fabrication and assembling. With regard to quality, attention is especially paid to the process monitoring and quality controls at the qualification step and during the routine production. Entirely different types of thermal UO 2 -PuO 2 fuel pellets, rods and assemblies have been manufactured for PWR and BWR operation. For these fabrications, some diagrams of the results with regard to the required technical specifications are presented. Special emphasis is placed on the occasional deviations of some finished products from the specifications and on the solutions applied to avoid such problems. Concerning the actual capacity of the mixed-oxide fuel fabrication plant, several limiting factors due to the nature of plutonium itself are discussed. Taking into account all these ambient limitations, a reference PWR mixed-oxide fuel output of nominally 18 t/a is obtained. The industrial feasibility of UO 2 -PuO 2 fuel fabrication has been thoroughly demonstrated by the present BELGONUCLEAIRE plant. The experience obtained has led to progressive improvements of the fabrication process and adaptation of the product controls in order to ensure the requested quality levels. (author)

  17. Calculation of parameters for inspection planning and evaluation: mixed-oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Reardon, P.T.; Mullen, M.F.

    1982-08-01

    As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities for mixed-oxide fuel fabrication facilities. There were four distinct efforts involved in this task. These were as follows: show the effect on a material balance verification of using two variables measurement methods in some strata; perform additional calculations for the reference facility described in STR-89; modify the INSPECT computer programs to be used as an after-inspection analysis tool, as well as a preinspection planning tool; provide written comments and explantations of text and graphs of the first draft of STR-89, Safeguards Considerations for Mixed-Oxide Fuel Element Fabrication Facilities, by W. Bahm, T. Shea, and D. Tolchenkov, System Studies Section, IAEA

  18. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    International Nuclear Information System (INIS)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques

  19. Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

    International Nuclear Information System (INIS)

    Brant Pinheiro, R.

    1993-01-01

    Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

  20. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    Energy Technology Data Exchange (ETDEWEB)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques.

  1. Case study application of the IAEA safeguards assessment methodology to a mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Swartz, J.; McDaniel, T.

    1981-01-01

    Science Applications, Inc. has prepared a case study illustrating the application of an assessment methodology to an international system for safeguarding mixed oxide (MOX) fuel fabrication facilities. This study is the second in a series of case studies which support an effort by the International Atomic Energy Agency (IAEA) and an international Consultant Group to develop a methodology for assessing the effectiveness of IAEA safeguards. 3 refs

  2. Synthesis and characterization of brannerite wasteforms for the immobilization of mixed oxide fuel residues

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, D.J.; Stennett, M.C.; Hyatt, N.C. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Sheffield, S1 3JD (United Kingdom)

    2016-07-01

    A possible method for the reduction of civil Pu stockpiles is the reuse of Pu in mixed oxide fuel (MOX). During MOX fuel production, residues unsuitable for further recycle will be produced. Due to their high actinide content MOX residues require immobilization within a robust host matrix. Although it is possible to immobilize actinides in vitreous wasteforms; ceramic phases, such as brannerite (UTi{sub 2}O{sub 6}), are attractive due to their high waste loading capacity and relative insolubility. A range of uranium brannerite, formulated Gd{sub x}U{sub 1-x}Ti{sub 2}O{sub 6}, were prepared using a mixed oxide route. Charge compensation of divalent and trivalent cations was expected to occur via the oxidation of U{sup 4+} to higher valence states (U{sup 5+} or U{sup 6+}). Gd{sup 3+} was added to act as a neutron absorber in the final Pu bearing wasteform. X-ray powder diffraction of synthesised specimens found that phase distribution was strongly affected by processing atmosphere (air or Ar). In all cases prototypical brannerite was formed accompanied by different secondary phases dependent on processing atmosphere. Microstructural analysis (SEM) of the sintered samples confirmed the results of the X-ray powder diffraction. The preliminary results presented here indicate that brannerite is a promising host matrix for mixed oxide fuel residues. (authors)

  3. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

    1999-01-01

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  4. Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi

    1985-01-01

    The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

  5. Reliability of fast reactor mixed-oxide fuel during operational transients

    International Nuclear Information System (INIS)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  6. Effect of a time varying power level in EBR-II on mixed-oxide fuel burnup

    International Nuclear Information System (INIS)

    Stone, I.Z.; Jost, J.W.; Baker, R.B.

    1979-01-01

    A refined prediction of burnup of mixed-oxide fuel in EBR-2 is compared with measured data. The calculation utilizes a time-varying power factor and results in a general improvement to previous calculations

  7. Status of plutonium recycle from mixed oxide fuel fabrication wastes (U,Pu)O2 facility activities

    International Nuclear Information System (INIS)

    Quesada, Calixto A.; Adelfang, Pablo; Greiner, G.; Orlando, Oscar S.; Mathot, Sergio R.

    1999-01-01

    Within the specific subject of mixed oxides corresponding to the Fuel Cycle activities performed at CNEA, the recovery of plutonium from wastes originated during tests and pre-fabrication stages is performed. (author)

  8. Behavior of mixed-oxide fuel elements during an overpower transient

    International Nuclear Information System (INIS)

    Tsai, H.; Shikakura, S.

    1993-01-01

    A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

  9. Studies on the dissolution of mixed oxide spent fuel from FBR

    International Nuclear Information System (INIS)

    Nemoto, Shin-ichi; Shibata, Atsuhiro; Shioura, Takao; Okamoto, Fumitoshi; Tanaka, Yasumasa

    1995-01-01

    At the Chemical Processing Facility(CPF) in the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation(PNC), since 1982 Laboratory scale hot experiments have been carried out on the development of reprocessing technology for FBR mixed oxide fuel. The spent fuel pins which have been used in out experiments were irradiated in Experimental Fast Reactor 'Joyo' Phenix (France) and DFR(UK). Burn-up of the fuel pins were 4,400-100,000 MWd/t. This paper Summarizes a dissolution study that have been performed to define the Key parameters affecting dissolution rate such as concentration of nitric acid, burn-up, and temperature. And this paper also discusses about the character of releasing 85 Kr in chopping and dissolution process, and about the amount of insoluble residue. (author)

  10. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    International Nuclear Information System (INIS)

    Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

    1999-01-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

  11. Detailed description of an SSAC at the facility level for mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-09-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in a mixed oxide fuel fabrication facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  12. Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

    International Nuclear Information System (INIS)

    Ikeuchi, Hirotomo; Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Washiya, Tadahiro

    2013-01-01

    An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm -3 ) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10 2 -10 3 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio. (author)

  13. Influence of oxygen-metal ratio on mixed-oxide fuel performance

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Leggett, R.D.

    1979-04-01

    The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

  14. Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

    International Nuclear Information System (INIS)

    Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

    2000-01-01

    The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

  15. Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.

    1997-01-01

    Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

  16. Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

    Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

  17. Autoradiographic measurement of Pu distribution in mixed-oxide nuclear fuel

    International Nuclear Information System (INIS)

    Green, D.R.; Rasmussen, D.E.; Gray, W.H.

    1976-09-01

    The autoradiographic method described was developed for rapid, economical determination of the Pu distribution and microhomogeneity in mixed oxide fuel. High Pu concentration regions of any size down to 13 microns in diameter can be reproducibly resolved using this method. The new method uses computerized scanning and analysis, and includes automatic self-calibration to virtually elimate variations resulting from photographic film and processing. The speed of this new method allows analysis of enough data to ensure statistical reliability of occurrence frequencies, even for sparse populations of Pu-rich regions with diameters greater than 60 microns. Determination of these occurrence frequencies is an important factor in controlling fuel quality to ensure safe, efficient operation in a Liquid Metal Fast Breeder Reactor

  18. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  19. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

    2010-01-31

    An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  20. Plutonium-containing aerosols found within containment enclosures in industrial mixed-oxide reactor fuel fabrication

    International Nuclear Information System (INIS)

    Newton, G.J.; Yeh, H.C.; Stanley, J.A.

    1977-01-01

    Mixed oxide (PuO 2 and UO 2 ) nuclear reactor fuel pellets are fabricated within safety enclosures at Babcock and Wilcox's Park Township site near Apollo, PA. Forty-two sample runs of plutonium-containing aerosols were taken from within glove boxes during routine industrial operations. A small, seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentration. Powder comminution and blending produced aerosols with lognormal size distributions characterized by activity median aerodynamic diameters (AMAD) of 1.89 +- 0.33 μm, sigma/sub g/ = 1.62 +- 0.09 and a gross alpha aerosol concentration range of 0.1 to 150 nCi/l. Slug pressing and grinding produced aerosols of AMAD = 3.08 +- 0.1 μm, sigma/sub g/ = 1.53 +- 0.01 and AMAD = 2.26 +- 0.16 μm, sigma/sub g/ = 1.68 +- 0.20, respectively. Gross alpha aerosol concentrations ranged from 3.4 to 450 nCi/l. Centerless grinding produced similar-sized aerosols but the gross alpha concentration ranged from 220 to 1690 nCi/l. In vitro solubility studies on selected LAPS samples in a lung fluid simulant indicate that plutonium mixed-oxide aerosols are more soluble than laboratory-produced plutonium aerosols

  1. Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

  2. General features of conceptual design for the pilot plant to manufacture fuel rods from mixed oxides

    International Nuclear Information System (INIS)

    Quesada, C.A.; Adelfang, P.; Esteban, A.; Aparicio, G.; Friedenthal, M.; Orlando, O.S.

    1987-01-01

    This paper conceptually describes: 1) the processes in the manufacturing lines; 2) the distribution of quality controls and glove boxes in manufacturing lines; 3) the Control and Radiological Safety Room; 4) the Dressing Room; 5) the requirements of the ventilation system. The plant will be located in the first floor of the Radiochemical Processes Laboratory building, occupying a surface of 600 m 2 . The necessary equipment for the following manufacturing lines will be provided: a) conversion from Pu(NO3)4 to PuO 2 (through Pu(III)oxalate); b) manufacture of homogeneous of mixed oxides of U and Pu; c) manufacture of (U,Pu)O 2 pellets; d) manufacture of fuel rods of mixed uranium and plutonium oxides. (Author)

  3. Fabrication of 0.5-inch diameter FBR mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Benecke, M.W.; McCord, R.B.

    1979-01-01

    Large diameter (0.535 inch) mixed oxide fuel pellets for Fast Breeder Reactor application were successfully fabricated by the cold-press-and-sinter technique. Enriched UO 2 , PuO 2 -UO 2 , and PuO 2 -ThO 2 compositions were fabricated into nominally 90% theoretical density pellets for the UO 2 and PuO 2 -UO 2 compositions, and 88% and 93% T.D. for the PuO 2 -ThO 2 compositions. Some processing adjustments were required to achieve satisfactory pellet quality and density. Furnace heating rate was reduced from 200 to 50 0 C/h for the organic binder burnout cycle for the large, 0.535-inch diameter pellets to eliminate pellet cracking during sintering. Additional preslugging steps and die wall lubrication during pressing were used to eliminate pressing cracks in the PuO 2 -ThO 2 pellets

  4. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    International Nuclear Information System (INIS)

    Mertyurek, Ugur; Gauld, Ian C.

    2016-01-01

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  5. EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

    International Nuclear Information System (INIS)

    Teague, Melissa C; Gorman, Brian P.; Miller, Brandon D; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel

  6. Calculational assessment of critical experiments with mixed-oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.; Funabashi, H.

    1987-01-01

    Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k/sub eff/ for organically moderated MOX fuel pin assemblies. Furthermore, the reactivity of an organic solution with a 32 vol % TBP/68 vol% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water

  7. Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

    1982-05-01

    Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

  8. Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.

    1987-01-01

    Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

  9. International safeguards for a modern MOX [mixed-oxide] fuel fabrication facility

    International Nuclear Information System (INIS)

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating σ/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials

  10. Oxygen-to-metal ratio control during fabrication of mixed oxide fast breeder reactor fuel pellets

    International Nuclear Information System (INIS)

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

    1979-05-01

    Oxygen-to-metal ratio (O/M) of mixed oxide fuel pellets can be controlled during fabrication by proper selection of binder (type and content) and sintering conditions. Sintering condition adjustments involved the passing of Ar--8% H 2 sintering gas across a cryostat ice bath controlled to temperatures ranging from -5 to -60 0 C to control as-sintered pellet O/M ratio. As-sintered fuel pellet O/M decreased with increasing Sterotex binder and PuO 2 concentrations, increasing sintering temperature, and decreasing sintering gas dew point. Approximate relationships between Sterotex binder level and O/M were established for PuO 2 --UO 2 and PuO 2 --ThO 2 fuels. O/M was relatively insensitive to Carbowax binder concentration. Several methods of increasing O/M using post-sintering pellet heat treatments were demonstrated, with the most reliable being a two-step process of first raising the O/M to 2.00 (stoichiometric) at 650 0 C in Ar--8% H 2 bubbled through H 2 O, followed by hydrogen reduction to specification O/M in oxygen-gettered Ar-8% H 2 at temperatures ranging from 1200 to 1690 0 C

  11. Waste management state-of-the-art review for mixed-oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Woodsum, H.C.; Goodman, J.

    1977-11-01

    This report provides a state-of-the-art review of the waste management for mixed-oxide (MOX) fuel fabrication facilities. The intent of this report is to focus on those processes and regulatory issues which have a direct bearing on existing and anticipated future management of transuranic (TRU) wastes from a commercial MOX fuel fabrication faciity. Recent government agency actions are reviewed with regard to their impact on existing and projected waste management regulations; and it is concluded that acceleration in the development of regulations, standards, and criteria is one of the most important factors in the implementation of improved MOX plant waste management techniques. ERDA development programs pertaining to the management of TRU wastes have been reviewed and many promising methods for volume reduction of both solid and liquid wastes are discussed. For solid wastes, these methods include compaction, shredding and baling, combustion, acid digestion, and decontamination by electropolishing or by electrolytic treatment. For liquid wastes, treatment options include evaporation, drying, calcination, flocculation, ion exchange, filtration, reverse osmosis, combustion (of combustible organics), and bioprocessing. Based on this review, it is recommended that ERDA continue with its combustible solid waste volume reduction program and complete these development activities by 1979. Following this, a critical evaluation of solid waste volume reduction techniques should be made to select the most promising systems for a commercial MOX fuel facility

  12. International safeguards for a modern MOX (mixed-oxide) fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating sigma/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials.

  13. Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 1. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C. E.; Murphy, E. S.; Schneider, K J

    1979-01-01

    Detailed technology, safety and cost information are presented for the conceptual decommissioning of a reference small mixed oxide fuel fabrication plant. Alternate methods of decommissioning are described including immediate dismantlement, safe storage for a period of time followed by dismantlement and entombment. Safety analyses, both occupational and public, and cost evaluations were conducted for each mode.

  14. Summary of the physical chemical analyses of mixed oxide nuclear fuel as they might influence biological behavior and internal dose

    International Nuclear Information System (INIS)

    Eidson, A.F.; Mewhinney, J.A.

    1987-01-01

    Twelve representative materials that might be accidentally released during the fabrication of mixed-oxide nuclear fuel pellets were studied using x-ray diffraction, infrared spectroscopy, energy dispersive x-ray fluorescence, alpha spectroscopy and in vitro dissolution methods. The results are related to a postulated exposure accident and to inhalation experiments using laboratory animals. 19 refs., 5 figs., 19 tabs

  15. Thermal conductivity of beginning-of-life uranium-plutonium mixed oxide fuel for fast reactor (Interim report)

    International Nuclear Information System (INIS)

    Inoue, Masaki; Mizuno, Tomoyasu; Asaga, Takeo

    1997-11-01

    Thermal conductivity of uranium-plutonium mixed oxide fuel for fast reactor at beginning-of-life was correlated based on the recent results in order to apply to the fuel design and the fuel performance analysis. A number of experimental results of unirradiated fuel specimens were corrected from open literatures and PNC internal reports and examined for the database. In this work two porosity correction factors were needed for high density fuel and low density fuel (around the current Monju specification). The universal porosity correction factor was not determined in this work. In the next step, theoretical and analytical considerations should be taken into account. (J.P.N.)

  16. Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

    1991-02-01

    The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

  17. Measurement of PuO2 spot in the mixed-oxide fuels

    International Nuclear Information System (INIS)

    Yokosuka, Yoshifumi; Kaneda, Kenichiro; Ishikawa, Eiji; Nakajima, Katsuaki; Kashima, Sadamitsu

    1974-01-01

    The homogeneity of mixed-oxide fuel is evaluated with the maximum diameter of PuO 2 spots. PuO 2 spot distribution is influenced by fuel production process and raw materials used. Analytical measurement of the PuO 2 spot distribution was carried out. α-autoradiograph was taken with nitrocellulose film as a detector. The film was exposed for 15 and 30 minutes to the sample pellets, and etched in 10% NaOH for 2 minutes at 60 0 C. Measurement of PuO 2 spots was performed manually with a projector. PuO 2 spot distribution was measured. There is no difference of spot distribution owing to the location in a pellet. However some discrepancy exists in spot number and maximum diameter between surface and inner part. This phenomenon is explained by considering that the diffusion on surface is more effective than in inner part in sintering process. Three pellets were taken out from a lot at random, and α-autoradiograph was taken longitudinally with a pellet and transversely with the others. PuO 2 spot number counted from three autoradiographs was about 150/cm 2 , and the standard deviation was about 1.5. No difference was observed among them. Concerning the influence of mixing method on PuO 2 spot formation, three methods were examined; ball mill mixing (with and without lining) and Kneader mixing. Kneader mixing gave the worst results in maximum spot diameter and spot distribution range. (Tai, I.)

  18. Electron donating and acid-base properties of cerium oxide and its mixed oxides with alumina

    International Nuclear Information System (INIS)

    Sugunan, S.; Jalaja, J.M.

    1994-01-01

    The electron donating properties of cerium oxide activated at 300, 500 and 800 degC and of its mixed oxides with alumina were examined based on the adsorption of electron acceptors exhibiting different electron affinities. The surface acidity/basicity of the oxides was determined by titrimetry; the H 0,max values are given. The limit of electron transfer from the oxide surface lies within the region of 1.77 and 2.40 eV in terms of the electron affinity of the electron acceptor. Cerium oxide promotes the electron donor nature of alumina while leaving the limit of electron transfer unchanged. 2 tabs., 4 figs., 13 refs

  19. Off gas processing device for degreasing furnace for uranium/plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ueda, Masaya; Akasaka, Takayuki; Noura, Takeshi.

    1996-01-01

    A low melting ingredient capturing-cooling trap connected to a degreasing sintering furnace by way of sealed pipelines, a burning/decomposing device for decomposing high melting ingredient gases discharged from the cooling trap by burning them and a gas sucking means for forming the flow of off gases are contained in a glovebox, the inside pressure of which is kept negative. Since the degreasing sintering furnace for uranium/plutonium mixed oxide fuels is disposed outside of the glovebox, operation can be performed safely without greatly increasing the scale of the device, and the back flow of gases is prevented easily by keeping the pressure in the inside of the glovebox negative. Further, a heater is disposed at the midway of the sealed pipelines from the degreasing sintering furnace to the cooling trap, the temperature is kept high to prevent deposition of low melting ingredients to prevent clogging of the sealed pipelines. Further, a portion of the pipelines is made extensible in the axial direction to eliminate thermal stresses caused by temperature change thereby enabling to extend the life of the sealed pipelines. (N.H.)

  20. Portable calorimeter system for nondestructive assay of mixed-oxide fuels

    International Nuclear Information System (INIS)

    Roche, C.T.; Perry, R.B.; Lewis, R.N.; Jung, E.A.; Haumann, J.R.

    1978-04-01

    Calorimetric assay provides a precise, nondestructive method to determine sample Pu content based on the heat emitted by decaying radionuclides. This measurement, in combination with a gamma-spectrometer analysis of sample isotopic content, yields the total sample Pu mass. The technique is applicable to sealed containers and is essentially independent of sample matrix configuration and elemental composition. Conventional calorimeter designs employ large water-bath heat sinks and lack the portability needed by inspection personnel. The ANL air-chamber isothermal calorimeters are low-thermal-capacitance devices which eliminate the need for large constant-temperature heat sinks. These instruments are designed to use a feedback system that applies power to maintain the sample chamber at a constant electrical resistance and, therefore, at a constant temperature. The applied-power difference between a Pu-containing sample and a blank determines the radioactive-decay power. The operating characteristics of a calorimeter designed for assaying mixed-oxide powders, fuel pellets, and Pu-containing solutions are discussed. This device consists of the calorimeter, sample preheatr, and a microprocessor-controlled data-acquisition system. The small-sample device weighs 18 kg and has a measurement cycle of 20 min, with a precision of 0.1% at 10 mW. A 100-min gamma-ray measurement gives the specific power with a precision of better than 1% for samples containing 1 to 2 g of plutonium

  1. Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

    International Nuclear Information System (INIS)

    Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao

    2000-01-01

    Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

  2. O/M ratio measurement in pure and mixed oxide fuels - where are we now?

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, J.; Chidester, K.; Thompson, M. [Los Alamos National Lab., NM (United States)

    2001-07-01

    The scale-down in the US and Russian nuclear weapons stockpiles has produced a surplus of weapons grade plutonium and highly enriched uranium. The incorporation into mixed-oxide fuel (MOX) is one of the currently favored routes for surplus weapons-grade plutonium. The use of MOX as a nuclear reactor fuel is well established, particularly in Europe and Japan but not in the US. The primary purpose of this investigation was to evaluate existing analytical techniques for their applicability to O/M (oxygen-to-metal ratio) measurements of MOX derived from excess weapons plutonium. The second objective of this investigation was to bring up-to-date the literature on O/M measurement methods, which has not been undertaken in over 20 years. There are several classification schemes that can be used to organize O/M measurement methods. The most popular schemes are based on (a) whether the analysis is performed in solution (wet chemical) or on solid material (dry), and (b) whether the concentration of major constituents are analyzed directly (direct) or are inferred (indirect). Solid state coulometric titration is currently used extensively in studies of phase equilibria, defect chemistry, thermochemical measurement of oxides, including ferrites. Regardless of which indirect method is used (solid state coulometric titration or thermogravimetry), a primary, direct method will also be required for the establishment of the MO{sub 2} reference state, determination of method bias, and periodic calibration. It was recommended that the following direct method be adapted for this purpose: oxygen measurement by inert gas fusion/carbon reduction, and total U, Pu by controlled potential coulometry. In a table are listed the experimental values of accuracy for about 30 O/M methods. (A.C.)

  3. Effects of variations in fuel pellet composition and size on mixed-oxide fuel pin performance

    International Nuclear Information System (INIS)

    Makenas, B.J.; Jensen, B.W.; Baker, R.B.

    1980-10-01

    Experiments have been conducted which assess the effects on fuel pin performance of specific minor variations from nominal in both fuel pellet size and pellet composition. Such pellets are generally referred to in the literature as rogue pellets. The effect of these rogue pellets on fuel pin and reactor performance is shown to be minimal

  4. Managing plutonium in Britain. Current options[Mixed oxide nuclear fuels; Nuclear weapons

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This is the report of a two day meeting to discuss issues arising from the reprocessing of plutonium and production of mixed oxide nuclear fuels in Britain. It was held at Charney Manor, near Oxford, on June 25 and 26, 1998, and was attended by 35 participants, including government officials, scientists, policy analysts, representatives of interested NGO's, journalists, a Member of Parliament, and visiting representatives from the US and Irish governments. The topic of managing plutonium has been a consistent thread within ORG's work, and was the subject of one of our previous reports, CDR 12. This particular seminar arose out of discussions earlier in the year between Dr. Frank Barnaby and the Rt. Hon. Michael Meacher MP, Minister for the Environment. With important decisions about the management of plutonium in Britain pending, ORG undertook to hold a seminar at which all aspects of the subject could be aired. A number of on-going events formed the background to this initiative. The first was British Nuclear Fuels' [BNFL] application to the Environment Agency to commission a mixed oxide fuel [MOX] plant at Sellafield. The second was BNFL's application to vary radioactive discharge limits at Sellafield. Thirdly, a House of Lords Select Committee was in process of taking evidence, on the disposal of radioactive waste. Fourthly, the Royal Society, in a recent report entitled Management of Separated Plutonium, recommended that 'the Government should commission a comprehensive review... of the options for the management of plutonium'. Four formal presentations were made to the meeting, on the subjects of Britain's plutonium policy, commercial prospects for plutonium use, problems of plutonium accountancy, and the danger of nuclear terrorism, by experts from outside the nuclear industry. It was hoped that the industry's viewpoint would also be heard, and BNFL were invited to present a paper, but declined on the grounds that they

  5. Control of the optical properties of silicon and chromium mixed oxides deposited by reactive magnetron sputtering

    International Nuclear Information System (INIS)

    Vergara, L.; Galindo, R. Escobar; Martinez, R.; Sanchez, O.; Palacio, C.; Albella, J.M.

    2011-01-01

    The development of mixed-oxide thin films allows obtaining materials with better properties than those of the different binary oxides, which makes them suitable for a great number of applications in different fields, such as tribology, optics or microelectronics. In this paper we investigate the deposition of mixed chromium and silicon oxides deposited by reactive magnetron sputtering with a view to use them as optical coatings with an adjustable refractive index. These films have been characterized by means of Rutherford backscattering spectrometry, Auger electron spectroscopy, X-ray diffraction, scanning electron microscopy, Fourier-transform infrared spectroscopy and spectroscopic ellipsometry so as to determine how the deposition conditions influence the characteristics of the material. We have found that the deposition parameter whose influence determines the properties of the films to a greater extent is the amount of oxygen in the reactive sputtering gas.

  6. Nondestructive characterization of mixed oxide pellets in welded nuclear fuel pins by neutron radiography and gamma-autoradiography

    International Nuclear Information System (INIS)

    Panakkal, J.P.; Ghosh, J.K.; Roy, P.R.

    1989-01-01

    Nondestructive evaluation of nuclear fuel pellets after the welding of fuel pins plays a vital role in assuring a safe and reliable operation of reactors. Some of the important characteristics to be monitored in low plutonium enriched mixed oxide fuel pellets are plutonium enrichment, size of plutonium dioxide agglomerates, incorrect loading and geometric shape. Experiments were carried out at Bhabha Atomic Research Centre, Bombay on experimental fuel pins containing mixed oxide pellets of different geometry (solid and annular), of different plutonium enrichment (0-6 w% of plutonium dioxide) and containing PuO 2 agglomerates of size 125-2000 microns to evaluate these characteristics nondestructively. Neutron radiography of these fuel pins was carried out using a swimming pool type reactor 'APSARA'. Results of quantitative evaluation of the neutron radiographs and a simple model correlating neutron interaction probability and the optical density are presented. Gamma autoradiography of these fuel pins showed that these parameters could be evaluated with a few limitations. This paper presents the experimental details, quantitative analysis of the radiographs by microdensitometry and merits and demerits of neutron radiography and gamma autoradiography for nondestructive charcterisation of nuclear fuel pellets. (orig.)

  7. Influence of thermal treatments on the basic and catalytic properties of Mg,Al-mixed oxides derived from hydrotalcites

    Directory of Open Access Journals (Sweden)

    Bastiani R.

    2004-01-01

    Full Text Available This work studied the influence of calcination conditions on basic properties and catalytic performance of Mg,Al-mixed oxides derived from a hydrotalcite sample (Al/(Al+Mg=0.20. Various heating rates, calcination atmospheres and lengths of calcination at 723K were evaluated. TPD of CO2 and retroaldolization of diacetone alcohol (DAA were used to determine the basic properties of the mixed oxides. The basic site density determined by TPD of CO2 showed a better correlation with catalytic activity for acetone/citral aldol condensation than the relative basicity obtained from retroaldolization of DAA. Calcination atmosphere was the parameter that influenced most the basic and the catalytic properties of the Mg,Al-mixed oxides, with calcination under dry air being the best choice.

  8. Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO{sub 2} fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

  9. Fuel-cladding chemical interaction correlation for mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1986-10-01

    A revised wastage correlation was developed for FCCI with fabrication and operating parameters. The expansion of the data base to 305 data sets provided sufficient data to employ normal statistical techniques for calculation of confidence levels without unduly penalizing predictions. The correlation based on 316 SS cladding also adequately accounts for limited measured depths of interaction for fuel pins with D9 and HTq cladding

  10. Thorium/uranium mixed oxide nano-crystals: Synthesis, structural characterization and magnetic properties

    International Nuclear Information System (INIS)

    Hudry, Damien; Griveau, Jean-Christophe; Apostolidis, Christos; Colineau, Eric; Rasmussen, Gert; Walter, Olaf; Wang, Di; Venkata Sai Kiran Chakravadhaluna; Courtois, Eglantine; Kubel, Christian

    2014-01-01

    One of the primary aims of the actinide community within nano-science is to develop a good understanding similar to what is currently the case for stable elements. As a consequence, efficient, reliable and versatile synthesis techniques dedicated to the formation of new actinide-based nano-objects (e.g., nano-crystals) are necessary. Hence, a 'library' dedicated to the preparation of various actinide based nano-scale building blocks is currently being developed. Nano-scale building blocks with tunable sizes, shapes and compositions are of prime importance. So far, the non-aqueous synthesis method in highly coordinating organic media is the only approach which has demonstrated the capability to provide size and shape control of actinide-based nano-crystals (both for thorium and uranium, and recently extended to neptunium and plutonium). In this paper, we demonstrate that the non-aqueous approach is also well adapted to control the chemical composition of the nano-crystals obtained when mixing two different actinides. Indeed, the controlled hot co-injection of thorium acetylacetonate and uranyl acetate (together with additional capping agents) into benzyl ether can be used to synthesize thorium/uranium mixed oxide nano-crystals covering the full compositional spectrum. Additionally, we found that both size and shape are modified as a function of the thorium/uranium ratio. Finally, the magnetic properties of the different thorium/uranium mixed oxide nano-crystals were investigated. Contrary to several reports, we did not observe any ferromagnetic behavior. As a consequence, ferromagnetism cannot be described as a universal feature of nano-crystals of non-magnetic oxides as recently claimed in the literature. (authors)

  11. Licensing issues associated with the use of mixed-oxide fuel in U.S. commercial nuclear reactors

    International Nuclear Information System (INIS)

    Williams, D.L. Jr.

    1997-04-01

    On January 14, 1997, the Department of Energy, as part of its Record of Decision on the storage and disposition of surplus nuclear weapons materials, committed to pursue the use of excess weapons-usable plutonium in the fabrication of mixed-oxide (MOX) fuel for consumption in existing commercial nuclear power plants. Domestic use of MOX fuel has been deferred since the late 1970s, principally due to nuclear proliferation concerns. This report documents a review of past and present literature (i.e., correspondence, reports, etc.) on the domestic use of MOX fuel and provides discussion on the technical and regulatory issues that must be addressed by DOE (and the utility/consortia selected by DOE to effect the MOX fuel consumption strategy) in obtaining approval from the Nuclear Regulatory Commission to use MOX fuel in one or a group of existing commercial nuclear power plants

  12. Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

    Science.gov (United States)

    Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

    2018-01-01

    In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

  13. Mg-Fe-mixed oxides derived from layered double hydroxides: A study of the surface properties

    Directory of Open Access Journals (Sweden)

    Marinković-Nedučin Radmila P.

    2011-01-01

    Full Text Available The influence of surface properties on the selectivity of the synthesized catalysts was studied, considering that their selectivity towards particular hydrocarbons is crucial for their overall activity in the chosen Fischer- -Tropsch reaction. Magnesium- and iron-containing layered double hydroxides (LDH, with the general formula: [Mg1-xFex(OH2](CO3x/2?mH2O, x = = n(Fe/(n(Mg+n(Fe, synthesized with different Mg/Fe ratio and their thermally derived mixed oxides were investigated. Magnesium was chosen because of its basic properties, whereas iron was selected due to its well-known high Fischer-Tropsch activity, redox properties and the ability to form specific active sites in the layered LDH structure required for catalytic application. The thermally less stable multiphase system (synthesized outside the optimal single LDH phase range with additional Fe-phase, having a lower content of surface acid and base active sites, a lower surface area and smaller fraction of smaller mesopores, showed higher selectivity in the Fischer-Tropsch reaction. The results of this study imply that the metastability of derived multiphase oxides structure has a greater influence on the formation of specific catalyst surface sites than other investigated surface properties.

  14. Evaluation of existing United States' facilities for use as a mixed-oxide (MOX) fuel fabrication facility for plutonium disposition

    International Nuclear Information System (INIS)

    Beard, C.A.; Buksa, J.J.; Chidester, K.; Eaton, S.L.; Motley, F.E.; Siebe, D.A.

    1995-01-01

    A number of existing US facilities were evaluated for use as a mixed-oxide fuel fabrication facility for plutonium disposition. These facilities include the Fuels Material Examination Facility (FMEF) at Hanford, the Washington Power Supply Unit 1 (WNP-1) facility at Hanford, the Barnwell Nuclear Fuel Plant (BNFP) at Barnwell, SC, the Fuel Processing Facility (FPF) at Idaho National Engineering Laboratory (INEL), the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), and the P-reactor at the Savannah River Site (SRS). The study consisted of evaluating each facility in terms of available process space, available building support systems (i.e., HVAC, security systems, existing process equipment, etc.), available regional infrastructure (i.e., emergency response teams, protective force teams, available transportation routes, etc.), and ability to integrate the MOX fabrication process into the facility in an operationally-sound manner that requires a minimum amount of structural modifications

  15. Emission computer tomography on a Dodewaard mixed oxide fuel pin. Comparative PIE work with non-destructive and destructive techniques

    Energy Technology Data Exchange (ETDEWEB)

    Buurveld, H.A.; Dassel, G.

    1993-12-01

    A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

  16. Basic properties of the mixed oxides obtained by thermal decomposition of hydrotalcites containing different metallic compositions

    Energy Technology Data Exchange (ETDEWEB)

    Valente, J.S.; Figueras, F.; Gravelle, M.; Kumbhar, P.; Lopez, J.; Besse, J.P.

    2000-01-25

    Carbonated layered double hydroxides (LDHs) containing Al, Fe, or Cr in a Mg(OH){sub 2} matrix or Al dissolved in hydroxides of Mg, Cu, Ni, Co, or Zn are used as precursors of basic catalysts. Decarbonation is studied by thermal analysis. The average basic strength, evaluated by the decarbonation temperature, is related to the partial charge of oxygen in the LDHs obtained from the Sanderson theory of electronegativity. The enthalpy of adsorption of CO{sub 2} on the resulting mixed oxides is measured by calorimetry. A homogeneous surface is generally observed for CO{sub 2} adsorption, with initial heats of adsorption close to those reported for MgO. The number of sites determined by this method is proportional to the rate constants for {beta}-isophorone isomerization, suggesting that both techniques measure surface properties. The layered structure in which OH{sup {minus}} is the compensating anion can be re-formed by hydration. This process does not appreciably change the adsorption of CO{sub 2}; thus, oxygens and hydroxyls show similar basic strengths in this case.

  17. Production equipment development needs for a 700 metric ton/year light water reactor mixed oxide fuel manufacturing plant

    International Nuclear Information System (INIS)

    Blahnik, D.E.

    1977-09-01

    A literature search and survey of fuel suppliers was conducted to determine how much development of production equipment is needed for a 700 metric tons/y LWR mixed-oxide (UO 2 --PuO 2 ) fuel fabrication plant. Results indicate that moderate to major production equipment development is needed in the powder and pellet processing areas. The equipment in the rod and assembly processing areas need only minor development effort. Required equipment development for a 700 MT/y plant is not anticipated to delay startup of the plant. The development, whether major or minor, can be done well within the time frame for licensing and construction of the plant as long as conventional production equipment is used

  18. Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1994-10-01

    From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

  19. Charge distribution on plutonium-containing aerosols produced in mixed-oxide reactor fuel fabrication and the laboratory

    International Nuclear Information System (INIS)

    Yeh, H.C.; Newton, G.J.; Teague, S.V.

    1976-01-01

    The inhalation toxicity of potentially toxic aerosols may be affected by the electrostatic charge on the particles. Charge may influence the deposition site during inhalation and therefore its subsequent clearance and dose patterns. The electrostatic charge distributions on plutonium-containing aerosols were measured with a miniature, parallel plate, aerosol electrical mobility spectrometer. Two aerosols were studied: a laboratory-produced 238 PuO 2 aerosol (15.8 Ci/g) and a plutonium mixed-oxide aerosol (PU-MOX, natural UO 2 plus PuO 2 , 0.02 Ci/g) formed during industrial centerless grinding of mixed-oxide reactor fuel pellets. Plutonium-238 dioxide particles produced in the laboratory exhibited a small net positive charge within a few minutes after passing through a 85 Kr discharger due to alpha particle emission removal of valence electrons. PU-MOX aerosols produced during centerless grinding showed a charge distribution essentially in Boltzmann equilibrium. The gross alpha aerosol concentrations (960-1200 nCi/l) within the glove box were sufficient to provide high ion concentrations capable of discharging the charge induced by mechanical and/or nuclear decay processes

  20. Review of direct electrical heating experiments on irradiated mixed-oxide fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Bandyopadhyay, G.

    1982-01-01

    Results of approximately 50 out-of-reactor experiments that simulated various stages of a loss-of-flow event with irradiated fuel are presented. The tests, which utilized the direct electrical heating technique to simulate nuclear heating, were performed either on fuel segments with their original cladding intact or on fuel segments that were extruded into quartz tubes. The test results demonstrated that the macro- and microscopic fuel behavior was dependent on a number of variables including fuel heating rate, thermal history prior to a transient, the number of heating cycles, type of cladding (quartz vs stainless steel), and fuel burnup

  1. Plutonium separation by reduction stripping. Application to processing of mixed oxide (U,Pu)O2 fuel fabrication wastes

    International Nuclear Information System (INIS)

    Arnal, Thierry; Cousinou, Gerard; Ganivet, Michel.

    1978-11-01

    A procedure is described for separating plutonium from a uranium VI and plutonium IV mixture contained in an organic phase (tributyl phosphate diluted in dodecane). This separation is obtained by extracting the plutonium III using two organic reducers: hydrazine and paraminophenol. Paraminophenol has excellent reducing qualities, similar to those of ferrous sulphamate, but has the added advantage of not contaminating extracted plutonium. This procedure is currently used in processing production wastes from mixed oxide (U,Pu)O 2 fuels; the installation using this procedure is described in detail in this paper. Operating results show the remarkable efficiency of this procedure: the separated plutonium and uranium mass flows have been increased to 185 and 350 g.h -1 respectively; the uranium contains less than 0.1 ppm of plutonium on completion of the purification cycle [fr

  2. Dry recovery test of plutonium-uranium mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Kinugasa, Manabu; Kawamata, Kazuhiko; Kashima, Sadamitsu

    1981-01-01

    The oxidation conditions for pulverizing directly Pu-U mixed oxide pellets without mechanical crushing were examined to simplify the process and to reduce radiation exposure during the dry recovery of highly enriched Pu pellets. The specimens used were the Pusub(0.3) Usub(0.7) Osub(2-x) pellets with different density, which were sintered at 1650 deg C for 2 hours under an atmosphere of 5 % H 2 - N 2 . The oxidation experiment was carried out under several conditions. The oxidation products were examined by weight gain, X-ray diffraction, appearance pictures, SEM photographs and so on. From these studies, it can be concluded that the oxidation in NO 2 diluted with air was very powerful, but if only the coarse spalling of Pusub(0.3) Usub(0.7) O 2 sintered pellets is required, it is sufficient to oxidize them in air for 1 hr in a temperature range from 400 to 600 deg C. (Asami, T.)

  3. Production of mixed oxide fuel for fast reactor irradiation test by co-precipitation

    International Nuclear Information System (INIS)

    Todokoro, Akio; Masuda, Sumio; Naruki, Kaoru; Kaya, Akira; Koizumi, Masumichi

    1974-01-01

    Studies were made on the production of homogeneous mixed oxide by co-precipitation. Experiments were made on the effects of the addition rate of ammonia water, precipitation temperature, aging time of co-precipitate, and pH value. Plutonium refined by anion exchange was mixed with solution of uranium. The concentration of free acid in the mixed solution was adjusted to 1.5-2.0 M, and the total volume of the solution was made 3.01. The weight of Pu and U in the solution was 100g. The solution was kept at a definite temperature while being stirred. Concentrated ammonia solution was added to the solution at a definite rate. The precipitate thus formed was filtrated after aging, then dried for 24 hours at 100 +- 2 0 C. Dried co-precipitate was calcinated for 1 hr at 550 0 C. The reduction for 4 hours at 800 0 C gave the mixed powder of PuO 2 and UO 2 . After pressing, the powder was sintered for 2 hours at 1700 0 C. The shrinkage ratio decreased as the activity and tap density of the original powder increased. The activity determined by specific surface area increased as the rate of ammonia water addition increased, and as the precipitation temperature rose. Tap density was independent of the rate of addition of ammonia water. The activity of the powder increased and the tap density decreased as the aging time of precipitate increased. (Fukutomi, T.)

  4. Thermochemical data and its use in modeling chemical behavior in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Gibby, R.L.; Woodley, R.E.; Adamson, M.G.; Johnson, C.E.

    1979-01-01

    The status of US activities to obtain fuel chemistry data is reviewed. Analytical expressions addressing basic needs of all fuel chemistry models are presented. Fission product concentrations during irradiation, oxygen-to-metal (O/M) at beginning-of-life and at burnup, and the potential in fuel-cladding gap at burnup are described

  5. Thermal ramp rate effects on mixed-oxide fuel swelling/gas release

    International Nuclear Information System (INIS)

    Hinman, C.A.; Randklev, E.H.

    1979-01-01

    Macroscopic swelling behavior of PNL-10 was compared to that of PNL-2 fuel and it was found that the swelling-threshold behavior is similar for similar thermal conditions. Transient fission gas release for the PNL-10 fuel is very similar to that observed for the PNL-2 fuel for similar thermal conditions

  6. Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

    International Nuclear Information System (INIS)

    Trellue, Holly R.

    2004-01-01

    The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

  7. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  8. Performance of commercially produced mixed-oxide fuels in EBR-II

    International Nuclear Information System (INIS)

    Hales, J.W.; Lawrence, L.A.

    1980-11-01

    Commercially produced fuels for the Fast Flux Test Facility (FFTF) were irradiated in EBR-II under conditions of high cladding temperature (approx. 700 0 C) and low power (approx. 200 W/cm) to verify that manufacturing processes did not introduce variables which significantly affect general fuel performance. Four interim examinations and a terminal examination were completed to a peak burnup of 5.2 at. % to provide irradiation data pertaining to fuel restructuring and dimensional stability at low fuel temperature, fuel-cladding reactions at high cladding temperature and general fuel behavior. The examinations indicate completely satisfactory irradiation performance for low heat rates and high cladding temperatures to 5.2 at. % burnup

  9. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  10. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    International Nuclear Information System (INIS)

    Hellwig, Ch.; Kasemeyer, U.

    2001-01-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm 3 . The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  11. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hellwig, Ch.; Kasemeyer, U

    2001-03-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm{sup 3}. The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  12. Guidebook on quality control of mixed oxides and gadolinium bearing fuels for light water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    Under the coverage of an efficient quality assurance system, quality control in nuclear fuel fabrication is an essential element to assure the reliable performance of all its components in service. Incentives to increase fuel performance, by extending reactor cycles or achieving higher burnups and, in some countries to use recycled plutonium in light water reactors (LWRs) necessitated the development of new types of fuels. In the first case, due to higher uranium enrichments, a burnable neutron absorber was integrated to the fuel pellets. Gadolinia was found to form a solid solution with Uranium dioxide and, to present a burnup rate which matches fissile uranium depletion. (U,Gd)O 2 fuels which have been successfully used since the seventies, in boiling water reactors have more recently found an increased utilization, in pressurized water reactors. This amply justifies the publication of this TECDOC to encourage authorities, designers and manufacturers of these types of fuel to establish a more uniform, adapted and effective system of control, thus promoting improved materials reliability and good performance in advanced fuel for light water reactors. The Guidebook is subdivided into four chapters written by different authors. A separate abstract was prepared for each of these chapters. Refs, figs and tabs

  13. Radiological safety aspects in the fabrication of mixed oxide fuel elements

    International Nuclear Information System (INIS)

    Krishnamurthi, T.N.; Janardhanan, S.; Soman, S.D.

    1981-01-01

    The problems of radiological safety in the fabrication of (U, Pu)O 2 fuel assemblies for fast reactors utilising high exposure plutonium are discussed. Derived working limits for plutonium as a function of the burn-up of RAPS (Rajasthan Atomic Power Station) fuel, external gamma and neutron exposures from feed product batches, finished fuel pins and assemblies are presented. Shielding requirements for the various glove box operations are also indicated. In general, high exposure plutonium handling calls for remote fabrication and automation at various stages would play a key role in minimising exposures to personnel in a large production plant. (author)

  14. Quality assurance program for surveillance of fast reactor mixed oxide fuel analytical chemistry

    International Nuclear Information System (INIS)

    Rein, J.E.; Zeigler, R.K.; Waterbury, G.R.; McClung, W.E.; Praetorius, P.R.; Delvin, W.L.

    1976-01-01

    An effective quality assurance program for the chemical analysis of nuclear fuel is essential to assure that the fuel will meet the strict chemical specifications required for optimum reactor performance. Such a program has been in operation since 1972 for the fuels manufactured for the Fast Flux Test Facility. This program, through the use of common quality control and calibration standards, has consistently provided high levels of agreement among laboratories in all areas of analysis. The paper presented gives a summary of the chemical specifications for the fuel and source material, an outline of the requirements for laboratory qualifications and the preparation of calibration and quality control materials, general administration details of the plan, and examples where the program has been useful in solving laboratory problems

  15. Tunable catalytic properties of bi-functional mixed oxides in ethanol conversion to high value compounds

    Energy Technology Data Exchange (ETDEWEB)

    Ramasamy, Karthikeyan K.; Gray, Michel J.; Job, Heather M.; Smith, Colin D.; Wang, Yong

    2016-04-10

    tA highly versatile ethanol conversion process to selectively generate high value compounds is pre-sented here. By changing the reaction temperature, ethanol can be selectively converted to >C2alcohols/oxygenates or phenolic compounds over hydrotalcite derived bi-functional MgO–Al2O3cata-lyst via complex cascade mechanism. Reaction temperature plays a role in whether aldol condensationor the acetone formation is the path taken in changing the product composition. This article containsthe catalytic activity comparison between the mono-functional and physical mixture counterpart to thehydrotalcite derived mixed oxides and the detailed discussion on the reaction mechanisms.

  16. Evaluation of the characteristics of high burnup and high plutonium content mixed oxide (MOX) fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed from JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2011 are destructive post irradiation examination (D-PIE) such as EPMA and SIMS at CEA (Commissariat a l'Enegie Atomique) facility. Cadarache and PIE data analysis. In the frame of irradiation test of high plutonium content MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility (SCK/CEN: Studiecentrum voor Kernenergie/Centre d'Etude l'Energie Nucleaire) in Belgium. The activities done in JFY 2011 are non-destructive post irradiation examination (ND-PIE) and D-PIE and PIE data analysis. In this report the results of EPMA and SIMS with high burnup irradiation test and the result of gamma spectrometry measurement which can give FP gas release rate are reported. (author)

  17. Fuel-pellet-fabrication experience using direct-denitration-recycle-PuO2-coprecipitated mixed oxide

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1980-01-01

    The fuel pellet fabrication experience described in this paper involved three different feed powders: coprecipitated PuO 2 -UO 2 which was flash calcined in a fluidized bed; co-direct denitrated PuO 2 -UO 2 ; and direct denitrated LWR recycle PuO 2 which was mechanically blended with natural UO 2 . The objectives of this paper are twofold; first, to demonstrate that acceptable quality fuel pellets were fabricated using feed powders manufactured by processes other than the conventional oxalate process; and second, to highlight some pellet fabrication difficulties experienced with the direct denitration LWR recycle PuO 2 feed material, which did not produce acceptable pellets. The direct denitration LWR recycle PuO 2 was available as a by-product and was not specifically produced for use in fuel pellet fabrication. Nevertheless, its characteristics and pellet fabrication behavior serve to re-emphasize the importance of continued process development involving both powder suppliers and fuel fabricators to close the fuel cycle in the future

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

    International Nuclear Information System (INIS)

    Matzke, H.

    1987-01-01

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

  19. Field characterization of plutonium aerosols in mixed-oxide fuel fabrication

    International Nuclear Information System (INIS)

    Newton, G.J.; Teague, S.V.; Yeh, H.C.

    1976-01-01

    Nuclear reactor fuel pellets of PuO 2 and UO 2 are fabricated within safety enclosures at Babcock and Wilcox's Parker Township Site near Apolla, Pa. Nineteen sample runs were taken from within glove boxes of aerosols formed during powder comminution and blending. Eight sampling runs were also taken of a centerless grinding operation during routine industrial operations. A small seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentrations. The potential toxicity of inhaled plutonium originating in the nuclear fuel cycle following accidental releases of these aerosols and possible inhalation by industrial workers is considered

  20. Analysis of Differences in Void Coefficient Predictions for Mixed-Oxide-Fueled Tight-Pitch Light Water Reactor Cells

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Shiroya, Seiji; Kanda, Keiji; Cathalau, Stephane; Carre, Franck-Olivier; Aizawa, Otohiko; Takeda, Toshikazu

    2000-01-01

    Analysis of the benchmark problems on the void coefficient of mixed-oxide (MOX)-fueled tight-pitch cells has been performed using the Japanese SRAC code system with the JENDL-3.2 library and the French APOLLO-2 code with the CEA93 library based on JEF-2.2. The benchmark problems have been specified to investigate the physical phenomena occurring during the progressive voidage of MOX-fueled tight-pitch lattices, such as high conversion light water reactor lattices, and to evaluate the impact of nuclear data and calculational methods. Despite the most recently compiled nuclear data libraries and the sophisticated calculation schemes employed in both code systems, the k ∞ and void reactivity values obtained by the two code systems show considerable discrepancy especially in the highly voided state. The discrepancy of k ∞ values shows an obvious dependence on void fraction and also has been shown to be sensitive to the isotopic composition of plutonium. The observed discrepancies are analyzed by being decomposed into contributing isotopes and reactions and have been shown to be caused by a complicated balance of both negative and positive components, which are mainly attributable to differences in a limited number of isotopes including 239 Pu, 241 Pu, 16 O, and stainless steel

  1. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

    2011-01-13

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

    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

  3. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

  4. The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

    International Nuclear Information System (INIS)

    Mattera, C.

    2003-01-01

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

  5. Design information questionnaire for a model mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Glancy, J.E.

    1976-05-01

    The model fuel plant is based on the proposed Westinghouse Anderson, S.C., plant and is typical of plants that will be constructed and operated in 1980 to 1990. A number of plant systems and procedures are uncertain, and in these cases judgment was used in describing relevant parameters in order to provide a complete model on which to design an inspection plan. The model plant does not, therefore, strictly represent any planned facility nor does it strictly represent the ideas of Westinghouse on plant design and material accountability. This report is divided into two sections. The first section is the IAEA Design Information Questionnaire form that contains an outline of all information requested. The second section is a complete listing of design information

  6. Generation of floor response spectra for mixed-oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    Arthur, D.F.; Murray, R.C.; Tokarz, F.J.

    1975-01-01

    Floor or amplified response spectra are generally used as input motion for seismic analysis of critical equipment and piping in nuclear power plants and related facilities. The floor spectra are normally the result of a time-history calculation of building response to ground shaking. However, alternate approximate methods have been suggested by both Kapur and Biggs. As part of a study for the Nuclear Regulatory Commission horizontal floor response spectra were generated and compared by all three methods. The dynamic analyses were performed on a model of the Westinghouse Recycle Fuels Plant Manufacturing Building (MOFFP). Input to the time-history calculations was a synthesized accelerogram whose response spectrum is similar to that in Regulatory Guide 1.60. The response spectrum of the synthetic ground motion was used as input to the Kapur and Biggs methods. Calculations were performed for both hard (3500 fps) and soft (1500 fps) foundation soils. Results of comparison of the three methods indicate that although the approximate methods could easily be made acceptable from a safety standpoint, they would be overly conservative. The time-history method will yield floor spectra which are less uncertain and less conservative for a relatively modest additional effort. (auth)

  7. Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

    International Nuclear Information System (INIS)

    Nithyadevi, Rajan; Thilagam, L.; Karthikeyan, R.; Pal, Usha

    2016-01-01

    Highlights: • Use of advanced computational code – DRAGON-5 using advanced self shielding model USS. • Testing the capability of DRAGON-5 code for the analysis of light water reactor system. • Wide variety of fuels LEU, MOX and spent fuel have been analyzed. • Parameters such as k ∞ , one, few and multi-group macroscopic cross-sections and fluxes were calculated. • Suitability of deterministic methodology employed in DRAGON-5 code is demonstrated for LWR. - Abstract: Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k ∞ ) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo

  8. Mixed oxides obtained from Co and Mn containing layered double hydroxides: Preparation, characterization, and catalytic properties

    International Nuclear Information System (INIS)

    Kovanda, Frantisek; Rojka, Tomas; Dobesova, Jana; Machovic, Vladimir; Bezdicka, Petr; Obalova, Lucie; Jiratova, Kveta; Grygar, Tomas

    2006-01-01

    Co-Mn-Al layered double hydroxides (LDHs) with various Co:Mn:Al molar ratios (4:2:0, 4:1.5:0.5, 4:1:1, 4:0.5:1.5, and 4:0:2) were prepared and characterized. Magnesium containing LDHs Co-Mg-Mn (2:2:2), Co-Mg-Mn-Al (2:2:1:1), and Co-Mg-Al (2:2:2) were also studied. Thermal decomposition of prepared LDHs and formation of related mixed oxides were studied using high-temperature X-ray powder diffraction and thermal analysis. The thermal decomposition of Mg-free LDHs starts by their partial dehydration accompanied by shrinkage of the lattice parameter c from ca. 0.76 to 0.66 nm. The dehydration temperature of the Co-Mn-Al LDHs decreases with increasing Mn content from 180 deg. C in Co-Al sample to 120 deg. C in sample with Co:Mn:Al molar ratio of 4:1.5:0.5. A subsequent step is a complete decomposition of the layered structure to nanocrystalline spinel, the complete dehydration, and finally decarbonation of the mixed oxide phase. Spinel-type oxides were the primary crystallization products. Mg-containing primary spinels had practically empty tetrahedral cationic sites. A dramatic increase of the spinel cell size upon heating and analysis by Raman spectroscopy revealed a segregation of Co-rich spinel in Co-Mn and Co-Mn-Al specimens. In calcination products obtained at 500 deg. C, the spinel mean coherence length was 5-10 nm, and the total content of the X-ray diffraction crystalline portion was 50-90%. These calcination products were tested as catalysts in the total oxidation of ethanol and decomposition of N 2 O. The catalytic activity in ethanol combustion was enhanced by increasing (Co+Mn) content while an optimum content of reducible components was necessary for high activity in N 2 O decomposition, where the highest conversions were found for calcined Co-Mn-Al sample with Co:Mn:Al molar ratio of 4:1:1

  9. Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

    International Nuclear Information System (INIS)

    Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

    1999-08-01

    The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

  10. Angular-resolution and material-characterization measurements for a dual-particle imaging system with mixed-oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Poitrasson-Rivière, Alexis, E-mail: alexispr@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Polack, J. Kyle; Hamel, Michael C.; Klemm, Dietrich D.; Ito, Kai; McSpaden, Alexander T.; Flaska, Marek; Clarke, Shaun D.; Pozzi, Sara A. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Tomanin, Alice [Lainsa-Italia S.R.L., Via E. Fermi 2749, 21027 Ispra, VA (Italy); Peerani, Paolo [European Commission, Joint Research Centre, Institute for Transuranium Elements, 21027 Ispra, VA (Italy)

    2015-10-11

    A dual-particle imaging (DPI) system, capable of simultaneously imaging fast neutrons and gamma rays, has been operated in the presence of mixed-oxide (MOX) fuel to assess the system's angular resolution and material-characterization capabilities. The detection principle is based on the scattering physics of neutrons (elastic scattering) and gamma rays (Compton scattering) in organic and inorganic scintillators. The detection system is designed as a combination of a two-plane Compton camera and a neutron-scatter camera. The front plane consists of EJ-309 liquid scintillators and the back plane consists of interleaved EJ-309 and NaI(Tl) scintillators. MCNPX-PoliMi was used to optimize the geometry of the system and the resulting prototype was built and tested using a Cf-252 source as an SNM surrogate. A software package was developed to acquire and process data in real time. The software was used for a measurement campaign to assess the angular resolution of the imaging system with MOX samples. Measurements of two MOX canisters of similar isotopics and intensity were performed for 6 different canister separations (from 5° to 30°, corresponding to distances of 21 cm and 131 cm, respectively). The measurements yielded a minimum separation of 20° at 2.5 m (86-cm separation) required to see 2 separate hot spots. Additionally, the results displayed good agreement with MCNPX-PoliMi simulations. These results indicate an angular resolution between 15° and 20°, given the 5° step. Coupled with its large field of view, and its capability to differentiate between spontaneous fission and (α,n) sources, the DPI system shows its potential for nuclear-nonproliferation applications.

  11. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  12. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 3

    International Nuclear Information System (INIS)

    1976-08-01

    An assessment is presented of the health, safety and environmental effects of the entire light water reactor fuel cycle, considering the comparative effects of three major alternatives: no recycle, recycle of uranium only, and recycle of both uranium and plutonium. The assessment covers the period from 1975 through the year 2000 and includes the cumulative effects for the entire period as well as projections for specific years. Topics discussed include: the light water reactor with plutonium recycle; mixed oxide fuel fabrication; reprocessing plant operations; supporting uranium fuel cycle; transportation of radioactive materials; radioactive waste management; storage of plutonium; radiological health assessment; extended spent fuel storage; and blending of plutonium and uranium at reprocessing plants

  13. Measurement of neutron energy spectra of PuO[sub 2]-UO[sub 2] mixed oxide fuel and penetrated through surrounding lead-acryl shield

    Energy Technology Data Exchange (ETDEWEB)

    Nakao, Noriaki; Tsujimura, Norio; Nakamura, Takashi (Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center); Momose, Takumaro; Ninomiya, Kazushige; Ishiguro; Hideharu

    1993-12-01

    The energy spectra of neutrons emitted from an aluminum can containing PuO[sub 2]-UO[sub 2] mixed oxide fuel and penetrated through a 35mm thick lead-acryl shield surrounding the can, were measured with the NE-213 organic liquid scintillator, the proton recoil proportional counter and the multi-moderator [sup 3]He spectrometer (Bonner Ball). The measured results were compared with the results calculated by the MORSE-CG Monte Carlo code on the basis of source neutron yields obtained by the ORIGEN-2 code and the source energy spectrum cited from the reference data. The agreement between these two was pretty good. The dose equivalents were then calculated from thus-obtained energy spectra and the flux-to-dose conversion factor and showed good agreement with the data measured with the neutron dose-equivalent counters (rem counters). Since the published data on energy spectrum of mixed oxide fuel are very scarce, these results can be useful as basic data for shielding design study and radiation control of nuclear fuel facilities. (author).

  14. Properties of large Li ion cells using a nickel based mixed oxide

    Science.gov (United States)

    Broussely, M.; Blanchard, Ph; Biensan, Ph; Planchat, J. P.; Nechev, K.; Staniewicz, R. J.

    The possible use of LiNiO 2 similar to LiCoO 2, as a positive material in rechargeable lithium batteries was recognized 20 years ago and starting 10 years later, many research studies led to material improvement through substitution of some of the nickel ions by other metallic ions. These modifications improve the thermal stability at high charge level or overcharge, as well as cycling and storage properties. Commercial material is now available at large industrial scale, which allows its use in big "industrial" Li ion batteries. Using low cost raw material (Ni), it is expected to be cost competitive with the manganese based systems usually mentioned as low cost on the total cell $/Wh basis. Providing higher energy density, and demonstrating excellent behavior on storage and extended cycle life, LiNiO 2 has definite advantages over the manganese system. Thanks to their properties, these batteries have demonstrated their ability to be used in lot of applications, either for transportation or standby. Their light weight makes them attractive for powering satellites. Although safety improvements are always desirable for all non-aqueous batteries using flammable organic electrolytes, suitable battery designs allow the systems to reach the acceptable level of safety required by many users. Beside the largely distributed lead acid and nickel cadmium batteries, Li ion will found its place in the "industrial batteries" market, in a proportion directly linked to its future cost reduction.

  15. Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors

    International Nuclear Information System (INIS)

    Bertini, H.W.

    1977-01-01

    Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO

  16. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Vallet, V.

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

  17. Physico-Chemical Properties of MgGa Mixed Oxides and Reconstructed Layered Double Hydroxides and Their Performance in Aldol Condensation of Furfural and Acetone

    Directory of Open Access Journals (Sweden)

    Oleg Kikhtyanin

    2018-05-01

    Full Text Available MgGa layered double hydroxides (Mg/Ga = 2–4 were synthesized and used for the preparation of MgGa mixed oxides and reconstructed hydrotalcites. The properties of the prepared materials were examined by physico-chemical methods (XRD, TGA, NH3-TPD, CO2-TPD, SEM, and DRIFT and tested in aldol condensation of furfural and acetone. The as-prepared phase-pure MgGa samples possessed hydrotalcite structure, and their calcination resulted in mixed oxides with MgO structure with a small admixture phase characterized by a reflection at 2θ ≈ 36.0°. The interaction of MgGa mixed oxides with pure water resulted in reconstruction of the HTC structure already after 15 s of the rehydration with maximum crystallinity achieved after 60 s. TGA-MS experiments proved a substantial decrease in carbonates in all rehydrated samples compared with their as-prepared counterparts. This allowed suggesting presence of interlayer hydroxyls in the samples. Acido-basic properties of MgGa mixed oxides determined by TPD technique did not correlate with Mg/Ga ratio which was explained by the specific distribution of Ga atoms on the external surface of the samples. CO2-TPD method was also used to evaluate the basic properties of the reconstructed MgGa samples. In these experiments, an intensive peak at T = 450°C on CO2-TPD curve was attributed to the decomposition of carbonates newly formed by CO2 interaction with interlayer carbonates rather than to CO2 desorption from basic sites. Accordingly, CO2-TPD method quantitatively characterized the interlayer hydroxyls only indirectly. Furfural conversion on reconstructed MgGa materials was much larger compared with MgGa mixed oxides confirming that Brønsted basic sites in MgGa catalysts, like MgAl catalysts, were active in the reaction. Mg/Ga ratio in mixed oxides influenced product selectivity which was explained by the difference in textural properties of the samples. In contrast, Mg/Ga ratio in reconstructed catalysts had

  18. Physico-Chemical Properties of MgGa Mixed Oxides and Reconstructed Layered Double Hydroxides and Their Performance in Aldol Condensation of Furfural and Acetone.

    Science.gov (United States)

    Kikhtyanin, Oleg; Čapek, Libor; Tišler, Zdeněk; Velvarská, Romana; Panasewicz, Adriana; Diblíková, Petra; Kubička, David

    2018-01-01

    MgGa layered double hydroxides (Mg/Ga = 2-4) were synthesized and used for the preparation of MgGa mixed oxides and reconstructed hydrotalcites. The properties of the prepared materials were examined by physico-chemical methods (XRD, TGA, NH 3 -TPD, CO 2 -TPD, SEM, and DRIFT) and tested in aldol condensation of furfural and acetone. The as-prepared phase-pure MgGa samples possessed hydrotalcite structure, and their calcination resulted in mixed oxides with MgO structure with a small admixture phase characterized by a reflection at 2θ ≈ 36.0°. The interaction of MgGa mixed oxides with pure water resulted in reconstruction of the HTC structure already after 15 s of the rehydration with maximum crystallinity achieved after 60 s. TGA-MS experiments proved a substantial decrease in carbonates in all rehydrated samples compared with their as-prepared counterparts. This allowed suggesting presence of interlayer hydroxyls in the samples. Acido-basic properties of MgGa mixed oxides determined by TPD technique did not correlate with Mg/Ga ratio which was explained by the specific distribution of Ga atoms on the external surface of the samples. CO 2 -TPD method was also used to evaluate the basic properties of the reconstructed MgGa samples. In these experiments, an intensive peak at T = 450°C on CO 2 -TPD curve was attributed to the decomposition of carbonates newly formed by CO 2 interaction with interlayer carbonates rather than to CO 2 desorption from basic sites. Accordingly, CO 2 -TPD method quantitatively characterized the interlayer hydroxyls only indirectly. Furfural conversion on reconstructed MgGa materials was much larger compared with MgGa mixed oxides confirming that Brønsted basic sites in MgGa catalysts, like MgAl catalysts, were active in the reaction. Mg/Ga ratio in mixed oxides influenced product selectivity which was explained by the difference in textural properties of the samples. In contrast, Mg/Ga ratio in reconstructed catalysts had practically

  19. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 2: A survey of the accuracy of the Studsvik of America CMS codes

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-02-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. Thus, before performing any kind of calculation with MOx fuels, it is necessary to be able to establish the reliability and the accuracy of these Core Management System (CMS) codes. This report presents a quantitative analysis of the models used in the package. A qualitative presentation is realized in a coming report

  20. Effect of burnup on the response of stainless steel-clad mixed-oxide fuels to simulated thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Badyopadhyay, G.

    1981-01-01

    Direct electrical heating experiments were performed on irradiated fuel to study the fuel and cladding response as a function of burnup during a slow thermal transient. The results indicated that the nature and extent of the fuel and cladding behavior depended on the quantity of fission gas retained in the fuel. Fission-gas-driven fuel ejection occurred as the molten cladding flowed down the stack exposing bare, radially unrestrained fuel. The fuel dispersion occurred in the absence of molten fuel and the amount of fuel ejected increased with increasing burnup. 31 refs

  1. Effect of Preparation Method on Catalytic Properties of Co-Mn-Al Mixed Oxides for N2O Decomposition.

    Czech Academy of Sciences Publication Activity Database

    Klyushina, A.; Pacultová, K.; Karásková, K.; Jirátová, Květa; Ritz, M.; Fridrichová, D.; Volodorskaja, A.; Obalová, L.

    2016-01-01

    Roč. 425, DEC 15 (2016), s. 237-247 ISSN 1381-1169 R&D Projects: GA ČR GA14-13750S Institutional support: RVO:67985858 Keywords : Co-Mn-Al mixed oxide * N2O decomposition * preparation methods Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 4.211, year: 2016

  2. Attestation in self-propagating combustion approach of spinel AFe_2O_4 (A = Co, Mg and Mn) complexes bearing mixed oxidation states: Magnetostructural properties

    International Nuclear Information System (INIS)

    Bennet, J.; Tholkappiyan, R.; Vishista, K.; Jaya, N. Victor; Hamed, Fathalla

    2016-01-01

    Highlights: • Spinel type ferrite compounds AFe_2O_4 (A = Co, Mg and Mn) have been successfully prepared by self-propagating combustion method using glycine as fuel. • To investigate and confirms the presence of phases in the synthesized ferrite nanoparticles by XRD and FTIR analysis. • The formation of mixed oxidation state of cobalt (Co"2"+ and Co"3"+), iron (Fe"2"+ and Fe"3"+) and manganese (Mn"2"+ and Mn"3"+) ions were studied and confirmed from XPS analysis. • The magnetic properties of the synthesized ferrites were studied by VSM measurement. - Abstract: Spinel type nano-sized ferrite compounds AFe_2O_4 (A = Co, Mg and Mn) have been successfully prepared by self-propagating combustion method using glycine as fuel at 400 °C under air atmosphere for 4 h. The crystal structure, chemical composition, morphology and magnetic properties of the synthesized samples were characterized by X−ray diffraction, Fourier transform infrared spectroscopy, X−ray photoelectron spectroscopy, Energy dispersive X−ray, Scanning and Transmission electron microscopy and vibrating sample magnetometer. The chemical reaction and role of fuel on the nanoparticles formation were discussed. The XRD pattern of the synthesized samples shows the formation of pure phase with average crystallite size of 97, 57 and 98 nm from Scherrer formula and 86, 54 and 87 nm from Williamson and Hall (W–H) formula respectively. FTIR absorption spectra revealed that the presence of strong absorption peaks near 400–600 cm"−"1 corresponds to tetrahedral and octahedral complex of spinel ferrites. The relative concentrations of electronic states of elements such as cobalt (Co"2"+ and Co"3"+), iron (Fe"2"+ and Fe"3"+) and manganese (Mn"2"+ and Mn"3"+) oxidation states were studied from XPS and it is found that 55% of Fe ions are in Fe"2"+ state and the remaining is in Fe"3"+ state and thus the cationic distribution of Fe ions occurred in both tetrahedral and octahedral sites. SEM analysis

  3. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  4. Mixed-oxide (MOX) fuel performance benchmark. Summary of the results for the PRIMO MOX rod BD8

    International Nuclear Information System (INIS)

    Ott, L.J.; Sartori, E.; Costa, A.; ); Sobolev, V.; Lee, B-H.; Alekseev, P.N.; Shestopalov, A.A.; Mikityuk, K.O.; Fomichenko, P.A.; Shatrova, L.P.; Medvedev, A.V.; Bogatyr, S.M.; Khvostov, G.A.; Kuznetsov, V.I.; Stoenescu, R.; Chatwin, C.P.

    2009-01-01

    The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A major part of these activities includes benchmark studies. This report describes the results of the PRIMO rod BD8 benchmark exercise, the second benchmark by the TFRPD relative to MOX fuel behaviour. The corresponding PRIMO experimental data have been released, compiled and reviewed for the International Fuel Performance Experiments (IFPE) database. The observed ranges (as noted in the text) in the predicted thermal and FGR responses are reasonable given the variety and combination of thermal conductivity and FGR models employed by the benchmark participants with their respective fuel performance codes

  5. SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

    International Nuclear Information System (INIS)

    Baker, R.B.; Wilson, D.R.

    1986-04-01

    The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

  6. Results of the post-irradiation examination of a highly-rated mixed oxide fuel rod from the Mol 7B experiment

    International Nuclear Information System (INIS)

    Coquerelle, M.; Walker, C.T.; Whitlow, W.H.

    1980-01-01

    The experiment MOL 7B was carried out in a epithermal flux in the Belgian reactor BR2. The pin examined contained fuel of initial composition (Usub(0.7)Pusub(0.3))Osub(1.98). It had been irradiated to a maximum burn-up of 13.2 at% at a maximum linear power of 568Wcm -1 . The fuel was clad with coldworked stainless steel. Electron microprobe analysis indicated that a Cr 2 O 3 type oxide was the main constituent of the grey phases in the gap. A metallic phase on the fuel surface had apparently resulted from the mechanical compaction of fragments of cladding that had been depleted in chromium by oxidation. Thus the main components of the phase were iron and nickel. Chromium loss from the inner cladding surface was significant only in the upper regions of the pin. In pin sections that were metallographically examined sigma phase and carbides of the type M 23 C 6 were present at the grain boundaries of the cladding. Cladding corrosion was not an Arrhenius function of the cladding temperature: the amount of metal lost from the inner cladding surface decreased with rise in cladding temperature above 910 K. A contributor to metal loss was the mechanical detachment of fragments of cladding which reformed as a metallic layer on the surface of the fuel. Chromium depletion and sigma phase formation at grain boundaries lowered the cohesive forces between grains which were then mechanically detached. Chromium loss from grain boundaries is mainly the results of oxidation of the cladding by the mixed oxide fuel. Data are presented to support the view that the local average O/M of the fuel determined the rate of oxidation and consequently the extent of chromium depletion. Fuel-cladding mechanical interactions were weak in the upper regions of the pin where metal loss was small

  7. Study of reactions between fuel (mixed oxide (UPu)Osub(2-x)) and cladding (stainless-steel) in reactors: influence of iodine compounds

    International Nuclear Information System (INIS)

    Aubert, Michel.

    1976-03-01

    The influence of iodine compounds on the development of the oxide-cladding reaction was examined. The action of iodine, cesium and cesium iodide on type 316 stainless was determined in the presence or absence of uranium oxide or mixed uranium-plutonium oxide type fuel in a closed system, isothermal or with a temperature gradient. The study of the stainless steel iodine reactions was developed in particular. These experiments showed that cesium combines with uranium oxide to give cesium uranate Cs 2 U 2 O 7 ; it is not unreasonable to suppose that cesium urano-plutonate Cs 2 (U,Pu) 2 O 7 could be formed inside the pile. It was then shown that cesium iodide in the presence of sufficiently non-stoichiometric mixed oxide could contribute towards the degradation of the stainless steel cladding. Under these conditions the reaction is accompained by a transport of manganese, chromium and iron into the hot parts of the fuel by a Van-Arkel type mechanism. This might explain the presence of metallic precipitates in the fuel, but the role assigned to molybdenum iodide in the same phenomenon is considered unlikely. Finally it is proposed to deposit a thin layer of manganese metal on the inner surface of the cladding in order to minimize the action of fission products (CsI, Te) [fr

  8. Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 2. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C. E.; Murphy, E. S.; Schneider, K. J.

    1979-01-01

    Volume 2 contains appendixes on small MOX fuel fabrication facility description, site description, residual radionuclide inventory estimates, decommissioning, financing, radiation dose methodology, general considerations, packaging and shipping of radioactive materials, cost assessment, and safety (JRD)

  9. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    International Nuclear Information System (INIS)

    Henderson, J.M.; Wood, S.A.; Knight, D.D.

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

  10. Design of a mixed-oxide fuel assembly to be assessed as a lead test assembly in a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez, H.; Alonso, G.

    2001-01-01

    The open and the close cycle are the two alternatives to pursue during power generation. The reprocessing is a mature process that now shows a more competitive economic aspect, making it more attractive than ever. Mexico has not decided what to do with the existing and future depleted fuel assemblies that will be generated from the power operation, thus the direct disposal and the reprocessing are still being considered. To have enough arguments in one or the other alternatives it is necessary to make an assessment of both. This investigation focus in the MOX fuel design assuming that the reprocessing is the option to follow and looking for the lowest impact in power generation. The first step in a reprocessing program is to analyze the performance of four lead test assemblies (LTA's), thus in this investigation we design the corresponding MOX to be used as LTA's and assess their performance through one operational cycle. (author)

  11. The uranium and thorium separation in the chemical reprocessing of the irradiated fuel of thorium and uranium mixed oxides

    International Nuclear Information System (INIS)

    Oliveira, E.F. de.

    1984-09-01

    A bibliographic research has been carried out for reprocessing techniques of irradiated thorium fuel from nuclear reactors. The Thorex/Hoechst process has been specially considered to establish a method for reprocessing thorium-uranium fuel from PWR. After a series of cold tests performed in laboratory it was possible to set the behavior of several parameters affecting the Thorex/Hoechst process. Some comments and suggestions are presented for modifications in the process flosheet conditions. A discussion is carried out for operational conditions such as the aqueous to organic flow ratio the acidity of strip and scrub solutions in the process steps for thorium and uranium recovery. The operation diagrams have been constructed using equilibrium experimental data which correspond to conditions observed in laboratory. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-07-01

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

  13. Euratom experience in safeguarding reprocessing and thermal reactor mixed oxide fuel fabrication facilities within the European Community

    International Nuclear Information System (INIS)

    1978-11-01

    The legal basis and instruments for the application of safeguards in the European Community are described. Euratom safeguards apply throughout the fuel cycle starting at the ore stage. Euratom has had experience in the application of safeguards to small and medium size reprocessing and MOX fabrication plants. In reprocessing plants accountancy, containment and surveillance methods are applied and the plant is divided into three material balance areas. Similar procedures are applied at fabrication plants. Euratom inspectors apply their main verification activities at strategic points but have the right of access at any time to all places which contain nuclear material. Under the Euratom-IAEA Agreements 'Joint Teams' of Euratom and IAEA inspectors will operate together to minimise the burden on operators and to avoid duplication of effort while enabling both organisations to achieve their safeguards objectives

  14. Boiling water reactors with Uranium-Plutonium mixed oxide fuel. Report 1: Accuracy of the nuclide concentrations calculated by CASMO-4

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. These CMS (Core Management System) programs have been extensively compared with both measurements and reference codes. Nevertheless some data are proprietary in particular the comparison of the calculated nuclide concentrations versus experiments (because of the cost of this kind of experimental study). This is why this report describes such a comparative investigation carried out with a General Electric 7x7 BWR bundle. Unfortunately, since some core history parameters were unknown, a lot of hypotheses have been adopted. This invokes sometimes a significant discrepancy in the results without being able to determine the origin of the differences between calculations and experiments. Yet one can assess that, except for four nuclides - Plutonium-238, Curium-243, Curium-244 and Cesium-135 - for which the approximate power history (history effect) can be invoked, the accuracy of the calculated nuclide concentrations is rather good if one takes the numerous approximations into account

  15. Thermochemical Properties of the Lattice Oxygen in W,Mn-Containing Mixed Oxide Catalysts for the Oxidative Coupling of Methane

    Science.gov (United States)

    Lomonosov, V. I.; Gordienko, Yu. A.; Sinev, M. Yu.; Rogov, V. A.; Sadykov, V. A.

    2018-03-01

    Mixed NaWMn/SiO2 oxide, samples containing individual components (Na, W, Mn) and their double combinations (Na-W, Na-Mn, W-Mn) supported on silica were studied by temperature programmed reduction (TPR) and desorption (TPD), and heat flow calorimetry during their reoxidation with molecular oxygen in pulse mode. The NaWMn/SiO2 mixed oxide was shown to contain two different types of reactive lattice oxygen. The weakly-bonded oxygen can be reversibly released from the oxide in a flow of inert gas in the temperature range of 575‒900°C, while the strongly-bonded oxygen can be removed during the reduction of the sample with hydrogen at 700-900°C. The measured thermal effect of oxygen consumption for these two oxygen forms are 185 and 350 kJ/mol, respectively. The amount of oxygen removed at reduction ( 443 μmol/g) considerably exceeded the amount desorbed in an inert gas flow ( 56 μmol/g). The obtained results suggest that the reversible oxygen desorption is due to the redox process in which manganese ions are involved, while during the temperature programmed reduction, mainly oxygen bonded with tungsten is removed.

  16. Application of Sensitivity and Uncertainty Analysis Methods to a Validation Study for Weapons-Grade Mixed-Oxide Fuel

    International Nuclear Information System (INIS)

    Dunn, M.E.

    2001-01-01

    At the Oak Ridge National Laboratory (ORNL), sensitivity and uncertainty (S/U) analysis methods and a Generalized Linear Least-Squares Methodology (GLLSM) have been developed to quantitatively determine the similarity or lack thereof between critical benchmark experiments and an application of interest. The S/U and GLLSM methods provide a mathematical approach, which is less judgment based relative to traditional validation procedures, to assess system similarity and estimate the calculational bias and uncertainty for an application of interest. The objective of this paper is to gain experience with the S/U and GLLSM methods by revisiting a criticality safety evaluation and associated traditional validation for the shipment of weapons-grade (WG) MOX fuel in the MO-1 transportation package. In the original validation, critical experiments were selected based on a qualitative assessment of the MO-1 and MOX contents relative to the available experiments. Subsequently, traditional trending analyses were used to estimate the Δk bias and associated uncertainty. In this paper, the S/U and GLLSM procedures are used to re-evaluate the suite of critical experiments associated with the original MO-1 evaluation. Using the S/U procedures developed at ORNL, critical experiments that are similar to the undamaged and damaged MO-1 package are identified based on sensitivity and uncertainty analyses of the criticals and the MO-1 package configurations. Based on the trending analyses developed for the S/U and GLLSM procedures, the Δk bias and uncertainty for the most reactive MO-1 package configurations are estimated and used to calculate an upper subcritical limit (USL) for the MO-1 evaluation. The calculated bias and uncertainty from the S/U and GLLSM analyses lead to a calculational USL that supports the original validation study for the MO-1

  17. In situ studies of uranium-plutonium mixed oxides. Influence of composition on phase equilibria and thermodynamic properties

    International Nuclear Information System (INIS)

    Strach, Michal

    2015-01-01

    Due to their physical and chemical properties, mixed uranium-plutonium oxides are considered for fuel in 4. generation nuclear reactors. In this frame, complementary experimental studies are necessary to develop a better understanding of the phenomena that take place during fabrication and operation in the reactor. The focus of this work was to study the U-Pu-O phase diagram in a wide range of compositions and temperatures to ameliorate our knowledge of the phase equilibria in this system. Most of experiments were done using in situ X-ray diffraction at elevated temperatures. The control of the oxygen partial pressure during the treatments made it possible to change the oxygen stoichiometry of the sample, which gave us an opportunity to study rapidly different compositions and the processes involved. The experimental approach was coupled with thermodynamic modeling using the CALPHAD method, to precisely plan the experiments and interpret the obtained results. This approach enabled us to enhance the knowledge of phase equilibria in the U-Pu-O system. (author) [fr

  18. Catalytic properties of perovskite-type mixed oxides, La/sub 1-x/Sr/sub x/CoO/sub 3/

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, T; Misono, M; Yoneda, Y

    1982-02-01

    The reactivity and related properties of oxygen both in the bulk and on the surface has been investigated for perovskite-type mixed oxides (La/sub 1-x/Sr,sub x/CoO/sub 3/), in regard to the effects of Sr/sup 2 +/-substitution and calcination temperature. The reducibility and the readiness of oxygen desorption increased with the Sr/sup 2 +/- content, x, but the reoxidation became slower with x. These results have been explained on the basis of the change in the chemical potential of lattice oxygen. The diffusivity of oxygen in the bulk and the ability to activate the oxygen molecule also increased with x. Oxygen vacancies in the bulk and on the surface, which tended to increase with x, are likely responsible for these reactions. The release of oxygen from the bulk became more difficult as the calcination temperature increased, in conformity with the trend of the catalytic activity.

  19. Solvothermal synthesis, characterization and optical properties of ZnO, ZnO-MgO and ZnO-NiO, mixed oxide nanoparticles

    International Nuclear Information System (INIS)

    Aslani, Alireza; Arefi, Mohammad Reza; Babapoor, Aziz; Amiri, Asghar; Beyki-Shuraki, Khalil

    2011-01-01

    ZnO-MgO and ZnO-NiO mixed oxides nanoparticles were produced from a solution containing Zinc acetate, Mg and Ni nitrate by Solvothermal method. The calcination process of the ZnO-MgO and ZnO-NiO composites nanoparticles brought forth polycrystalline two-phase ZnO-MgO and ZnO-NiO nanoparticles of 40-80 nm in diameters. ZnO, MgO and NiO were crystallized into wuertzite and rock salt structures, respectively. The optical properties of ZnO-MgO and ZnO-NiO nanoparticles were obtained by solid state UV and solid state florescent. The XRD, SEM and Raman spectroscopies of these nanoparticles were analyzed.

  20. Solvothermal synthesis, characterization and optical properties of ZnO, ZnO-MgO and ZnO-NiO, mixed oxide nanoparticles

    Science.gov (United States)

    Aslani, Alireza; Arefi, Mohammad Reza; Babapoor, Aziz; Amiri, Asghar; Beyki-Shuraki, Khalil

    2011-03-01

    ZnO-MgO and ZnO-NiO mixed oxides nanoparticles were produced from a solution containing Zinc acetate, Mg and Ni nitrate by Solvothermal method. The calcination process of the ZnO-MgO and ZnO-NiO composites nanoparticles brought forth polycrystalline two-phase ZnO-MgO and ZnO-NiO nanoparticles of 40-80 nm in diameters. ZnO, MgO and NiO were crystallized into würtzite and rock salt structures, respectively. The optical properties of ZnO-MgO and ZnO-NiO nanoparticles were obtained by solid state UV and solid state florescent. The XRD, SEM and Raman spectroscopies of these nanoparticles were analyzed.

  1. Solvothermal synthesis, characterization and optical properties of ZnO, ZnO-MgO and ZnO-NiO, mixed oxide nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Aslani, Alireza, E-mail: a.aslani@vru.ac.ir [Department of Chemistry, Faculty of Science, Vali-E-Asr University of Rafsanjan, Rafsanjan, PO Box: 77176 (Iran, Islamic Republic of); Arefi, Mohammad Reza [Islamic Azad University, Yazd Brunch, Young researchers Club, Yazd (Iran, Islamic Republic of); Babapoor, Aziz [Department of Chemical Engineering, Ahar Branch, Islamic Azad University, Ahar, PO Box: 54516 (Iran, Islamic Republic of); Amiri, Asghar; Beyki-Shuraki, Khalil [Department of Chemistry, Faculty of Science, Vali-E-Asr University of Rafsanjan, Rafsanjan, PO Box: 77176 (Iran, Islamic Republic of)

    2011-03-15

    ZnO-MgO and ZnO-NiO mixed oxides nanoparticles were produced from a solution containing Zinc acetate, Mg and Ni nitrate by Solvothermal method. The calcination process of the ZnO-MgO and ZnO-NiO composites nanoparticles brought forth polycrystalline two-phase ZnO-MgO and ZnO-NiO nanoparticles of 40-80 nm in diameters. ZnO, MgO and NiO were crystallized into wuertzite and rock salt structures, respectively. The optical properties of ZnO-MgO and ZnO-NiO nanoparticles were obtained by solid state UV and solid state florescent. The XRD, SEM and Raman spectroscopies of these nanoparticles were analyzed.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    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)

  3. Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

    International Nuclear Information System (INIS)

    Strain, R.V.; Johnson, C.E.

    1978-05-01

    Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

  4. Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Tingey, Joel M.; Jones, Susan A.

    2005-01-01

    Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at

  5. Processing of FRG mixed oxide fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    International Nuclear Information System (INIS)

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Tischer, H.E.

    1980-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment section of the agreement, FRG fuel spheres were recently sent for processing in the Department of Energy sponsored cold pilot plant for High-Temperature Gas-Cooled Reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles. These particles were in turn crushed and burned to recover the fuel-bearing kernels for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated the applicability of the US HTGR fuel treatment flowsheet to FRG fuel processing. 10 figures

  6. Process for fabricating mixed-oxide powders

    International Nuclear Information System (INIS)

    Elmaleh, D.; Giraudel, A.

    1975-01-01

    A physical-chemical process for fabricating homogeneous powders suitable for sintering is described. It can be applied to the synthesis of all mixed oxides having mutually compatible and water soluble salts. As a specific example, the fabrication of lead titanate-zirconate powders used to make hot pressed ceramics is described. These ceramics show improved piezoelectric properties [fr

  7. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Primm III, RT

    2002-05-29

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

  8. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    International Nuclear Information System (INIS)

    Primm III, RT

    2002-01-01

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors

  9. Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

    Energy Technology Data Exchange (ETDEWEB)

    Strain, R V; Johnson, C E

    1978-05-01

    Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

  10. Reprocessing facility for spent fuel from LWR type reactors and mixed-oxide fuel fabrication plant in the Taxoeldern Forest near Wackersdorf, Bavaria (WAA) - first partial licence

    International Nuclear Information System (INIS)

    1985-01-01

    Full text of the first partial licence for the WAA, allowing erection of the following buildings or structures: External fence; guardhouse 1, i.e. the building and the ground connection system with lightning protection system, the fire alarm system and mobile fire-fighting systems; the fuel receiving station, including building and operation systems; excavation work for the main reprocessing building. (HP) [de

  11. The decomposition of mixed oxide Ag2Cu2O3: Structural features and the catalytic properties in CO and C2H4 oxidation

    Science.gov (United States)

    Svintsitskiy, Dmitry A.; Kardash, Tatyana Yu.; Slavinskaya, Elena M.; Stonkus, Olga A.; Koscheev, Sergei V.; Boronin, Andrei I.

    2018-01-01

    The mixed silver-copper oxide Ag2Cu2O3 with a paramelaconite crystal structure is a promising material for catalytic applications. The as-prepared sample of Ag2Cu2O3 consisted of brick-like particles extended along the [001] direction. A combination of physicochemical techniques such as TEM, XPS and XRD was applied to investigate the structural features of this mixed silver-copper oxide. The thermal stability of Ag2Cu2O3 was investigated using in situ XRD under different reaction conditions, including a catalytic CO + O2 mixture. The first step of Ag2Cu2O3 decomposition was accompanied by the appearance of ensembles consisting of silver nanoparticles with sizes of 5-15 nm. Silver nanoparticles were strongly oriented to each other and to the surface of the initial Ag2Cu2O3 bricks. Based on the XRD data, it was shown that the release of silver occurred along the a and b axes of the paramelaconite structure. Partial decomposition of Ag2Cu2O3 accompanied by the formation of silver nanoparticles was observed during prolonged air storage under ambient conditions. The high reactivity is discussed as a reason for spontaneous decomposition during Ag2Cu2O3 storage. The full decomposition of the mixed oxide into metallic silver and copper (II) oxide took place at temperatures higher than 300 °C regardless of the nature of the reaction medium (helium, air, CO + O2). Catalytic properties of partially and fully decomposed samples of mixed silver-copper oxide were measured in low-temperature CO oxidation and C2H4 epoxidation reactions.

  12. Attestation in self-propagating combustion approach of spinel AFe{sub 2}O{sub 4} (A = Co, Mg and Mn) complexes bearing mixed oxidation states: Magnetostructural properties

    Energy Technology Data Exchange (ETDEWEB)

    Bennet, J., E-mail: b.eenneett@gmail.com [Department of Physics, College of Engineering, Guindy, Anna University, Sardar Patel Road, Chennai,600025 (India); Tholkappiyan, R. [Department of Physics, College of Engineering, Guindy, Anna University, Sardar Patel Road, Chennai,600025 (India); Department of Physics, College of Science, UAE University, Al Ain 15551 (United Arab Emirates); Vishista, K.; Jaya, N. Victor [Department of Physics, College of Engineering, Guindy, Anna University, Sardar Patel Road, Chennai,600025 (India); Hamed, Fathalla [Department of Physics, College of Science, UAE University, Al Ain 15551 (United Arab Emirates)

    2016-10-15

    Highlights: • Spinel type ferrite compounds AFe{sub 2}O{sub 4} (A = Co, Mg and Mn) have been successfully prepared by self-propagating combustion method using glycine as fuel. • To investigate and confirms the presence of phases in the synthesized ferrite nanoparticles by XRD and FTIR analysis. • The formation of mixed oxidation state of cobalt (Co{sup 2+} and Co{sup 3+}), iron (Fe{sup 2+} and Fe{sup 3+}) and manganese (Mn{sup 2+} and Mn{sup 3+}) ions were studied and confirmed from XPS analysis. • The magnetic properties of the synthesized ferrites were studied by VSM measurement. - Abstract: Spinel type nano-sized ferrite compounds AFe{sub 2}O{sub 4} (A = Co, Mg and Mn) have been successfully prepared by self-propagating combustion method using glycine as fuel at 400 °C under air atmosphere for 4 h. The crystal structure, chemical composition, morphology and magnetic properties of the synthesized samples were characterized by X−ray diffraction, Fourier transform infrared spectroscopy, X−ray photoelectron spectroscopy, Energy dispersive X−ray, Scanning and Transmission electron microscopy and vibrating sample magnetometer. The chemical reaction and role of fuel on the nanoparticles formation were discussed. The XRD pattern of the synthesized samples shows the formation of pure phase with average crystallite size of 97, 57 and 98 nm from Scherrer formula and 86, 54 and 87 nm from Williamson and Hall (W–H) formula respectively. FTIR absorption spectra revealed that the presence of strong absorption peaks near 400–600 cm{sup −1} corresponds to tetrahedral and octahedral complex of spinel ferrites. The relative concentrations of electronic states of elements such as cobalt (Co{sup 2+} and Co{sup 3+}), iron (Fe{sup 2+} and Fe{sup 3+}) and manganese (Mn{sup 2+} and Mn{sup 3+}) oxidation states were studied from XPS and it is found that 55% of Fe ions are in Fe{sup 2+} state and the remaining is in Fe{sup 3+} state and thus the cationic distribution

  13. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  14. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  15. Influence of basic properties of Mg,Al-mixed oxides on their catalytic activity in knoevenagel condensation between benzaldehyde and phenylsulfonylacetonitrile

    Directory of Open Access Journals (Sweden)

    Caridad Noda Pérez

    2009-01-01

    Full Text Available The catalytic performance of Mg,Al-mixed oxides (MO20, MO25 and MO33 derived from hydrotalcites was evaluated in the Knoevenagel reaction between benzaldehyde and phenylsulfonylacetonitrile at 373 and 383 K. The best results were obtained for the sample MO20 that presented the highest basic sites density and external area and the smallest crystallite sizes. The relative amount of basic sites with weak to intermediate strength also played an important role on catalytic performance. By increasing the catalyst content from 1 to 5 wt.% at 383 K, a complete conversion of the reactants is attained, producing α-phenylsulfonylcinnamonitrile with a selectivity of 100%.

  16. Radiological safety aspects in the fabrication of mixed oxide fuel elements. [Derived working limits in air and water for plutonium, enriched uranium and their mixture

    Energy Technology Data Exchange (ETDEWEB)

    Krishnamurthi, T.N.; Janardhanan, S.; Soman, S.D. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

    The problems of radiological safety in the fabrication of (U, Pu)O/sub 2/ fuel assemblies for fast reactors utilising high exposure plutonium are discussed. Derived working limits for plutonium as a function of the burn-up of RAPS (Rajasthan Atomic Power Station) fuel, external gamma and neutron exposures from feed product batches, finished fuel pins and assemblies are presented. Shielding requirements for the various glove box operations are also indicated. In general, high exposure plutonium handling calls for remote fabrication and automation at various stages would play a key role in minimising exposures to personnel in a large production plant.

  17. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  18. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 2

    International Nuclear Information System (INIS)

    1976-08-01

    This environmental statement assesses the impacts of the implementation of plutonium recycle in the LWR industry. It is based on assumptions that are intended to reflect conservatively an acceptable level of the application of current technology. It is not intended to be a representation of the ''as low as reasonably achievable'' (ALARA) philosophy. This generic environmental statement discusses the anticipated effects of recycling plutonium in light water nuclear power reactors. It is based on about 30 years of experience with the element in the context of a projected light water nuclear power industry that is already substantial. A background perspective on plutonium, its safety, and its recycling as a reactor fuel is presented

  19. Influence of the physico-chemical properties of CeO{sub 2}-ZrO{sub 2} mixed oxides on the catalytic oxidation of NO to NO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Atribak, Idriss; Guillen-Hurtado, Noelia; Bueno-Lopez, Agustin [MCMA Group, Department of Inorganic Chemistry, Faculty of Sciences, University of Alicante, Carretera San Vicente del Raspeig s/n - 03690 San Vicente del Raspeig, Alicante (Spain); Garcia-Garcia, Avelina, E-mail: a.garcia@ua.es [MCMA Group, Department of Inorganic Chemistry, Faculty of Sciences, University of Alicante, Carretera San Vicente del Raspeig s/n - 03690 San Vicente del Raspeig, Alicante (Spain)

    2010-10-01

    Commercial and home-made Ce-Zr catalysts prepared by co-precipitation were characterised by XRD, Raman spectroscopy, N{sub 2} adsorption at -196 deg. C and XPS, and were tested for NO oxidation to NO{sub 2}. Among the different physico-chemical properties characterised, the surface composition seems to be the most relevant one in order to explain the NO oxidation capacity of these Ce-Zr catalysts. As a general trend, Ce-Zr catalysts with a cerium-rich surface, that is, high XPS-measured Ce/Zr atomic surface ratios, are more active than those with a Zr-enriched surface. The decrease in catalytic activity of the Ce-Zr mixed oxided upon calcinations at 800 deg. C with regard to 500 deg. C is mainly attributed to the decrease in Ce/Zr surface ratio, that is, to the surface segregation of Zr. The phase composition (cubic or t'' for Ce-rich compositions) seems not to be a direct effect on the catalytic activity for NO oxidation in the range of compositions tested. However, the formation of a proper solid solution prevents important surface segregation of Zr upon calcinations at high temperature. The effect of the BET surface area in the catalytic activity for NO oxidation of Ce-Zr mixed oxides is minor in comparison with the effect of the Ce/Zr surface ratio.

  20. Antimony (V) oxide adsorbed on a silica-zirconia mixed oxide obtained by the sol-gel processing method: preparation and acid properties

    International Nuclear Information System (INIS)

    Zaitseva, Galina; Gushikem, Yoshitaka

    2002-01-01

    The preparation, degree of dispersion, thermal stability and Lewis and Broensted acidity of antimony (V) oxide adsorbed on Si O 2 /Zr O 2 mixed oxide, previously prepared by the sol-gel processing method, are described herein. The samples Si O 2 /Zr O 2 /Sb 2 O 5 , with compositions (in wt %): (a) Zr= 8.1 and Sb= 6.3; (b) Zr= 14.9 and Sb= 11.4, were prepared. Scanning electron microscopy images connected to a X-ray energy dispersive spectrometer showed that both metals, Zr and Sb, were highly dispersed in the matrices. The X-ray diffraction patterns of Si O 2 /Zr O 2 /Sb 2 O 5 heated at different temperatures showed that, up to 1023 K, the matrix was amorphous. At 1273 K a crystalline phase of Zr O 2 (baddeleyite) was observed and raising the temperature to 1573 K, crystalline phases of Si O 2 (cristobalite) and presumably of Zr O 2 /Sb 2 O 5 were observed. Using pyridine as a molecular probe, Broensted acid sites on the Si O 2 /Zr O 2 /Sb 2 O 5 matrix, heat treated up to 523 K, were shown to be very stable. (author)

  1. Antimony (V Oxide Adsorbed on a Silica-Zirconia Mixed Oxide Obtained by the Sol-Gel Processing Method: Preparation and Acid Properties

    Directory of Open Access Journals (Sweden)

    Zaitseva Galina

    2002-01-01

    Full Text Available The preparation, degree of dispersion, thermal stability and Lewis and Brphinsted acidity of antimony (V oxide adsorbed on SiO2/ZrO2 mixed oxide, previously prepared by the sol-gel processing method, are described herein. The samples SiO2/ZrO2/Sb2O5, with compositions (in wt %: (a Zr= 8.1 and Sb= 6.3; (b Zr= 14.9 and Sb= 11.4, were prepared. Scanning electron microscopy images connected to a X-ray energy dispersive spectrometer showed that both metals, Zr and Sb, were highly dispersed in the matrices. The X-ray diffraction patterns of SiO2/ZrO2/Sb2O5 heated at different temperatures showed that, up to 1023 K, the matrix was amorphous. At 1273 K a crystalline phase of ZrO2 (baddeleyite was observed and raising the temperature to 1573 K, crystalline phases of SiO2 (cristobalite and presumably of ZrO2/Sb2O5 were observed. Using pyridine as a molecular probe, Brphinsted acid sites on the SiO2/ZrO2/Sb2O5 matrix, heat treated up to 523 K, were shown to be very stable.

  2. Fuel Property Blend Model

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wagnon, Scott J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zhang, Kuiwen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kukkadapu, Goutham [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-01-12

    The object of this project is to develop chemical models and associated correlations to predict the blending behavior of bio-derived fuels when mixed with conventional fuels like gasoline and diesel fuels.

  3. Analysis of the porosity distribution of mixed oxide pins

    International Nuclear Information System (INIS)

    Lieblich, M.; Lopez, J.

    1987-01-01

    In the frame of the Joint Irradiation Program IVO-FR2-Vg7 between the Centre of Nuclear Research of Karlsruhe (KfK), the irradiation of 30 mixed-oxide fuel rods in the FR2 experimental reactor was carried out. The pins were located in 10 single-walled NaK capsules. The behaviour of the fuel during its burnup was studied, mainly, the rest-porosity and cracking distribution in the pellet, partial densification, etc. In this work 3 pins from the capsule No. 165 were analyzed. The experimental results (pore and cracking profiles) were interpreted by the fuel rod code SATURN. (Author) 20 refs

  4. Reducibility of ceria-lanthana mixed oxides under temperature programmed hydrogen and inert gas flow conditions

    International Nuclear Information System (INIS)

    Bernal, S.; Blanco, G.; Cifredo, G.; Perez-Omil, J.A.; Pintado, J.M.; Rodriguez-Izquierdo, J.M.

    1997-01-01

    The present paper deals with the preparation and characterization of La/Ce mixed oxides, with La molar contents of 20, 36 and 57%. We carry out the study of the structural, textural and redox properties of the mixed oxides, comparing our results with those for pure ceria. For this aim we use temperature programmed reduction (TPR), temperature programmed desorption (TPD), nitrogen physisorption at 77 K, X-ray diffraction and high resolution electron microscopy. The mixed oxides are more easy to reduce in a flow of hydrogen than ceria. Moreover, in an inert gas flow they release oxygen in higher amounts and at lower temperatures than pure CeO 2 . The textural stability of the mixed oxides is also improved by incorporation of lanthana. All these properties make the ceria-lanthana mixed oxides interesting alternative candidates to substitute ceria in three-way catalyst formulations. (orig.)

  5. Grain growth kinetics in uranium-plutonium mixed oxides

    International Nuclear Information System (INIS)

    Sari, C.

    1986-01-01

    Grain growth rates were investigated in uranium-plutonium mixed oxide specimens with oxygen-to-metal ratios 1.97 and 2.0. The specimens in the form of cylindrical pellets were heated in a temperature gradient similar to that existing in a fast reactor. The results are in agreement with the cubic rate law. The mean grain size D(μm) after annealing for time t (min) is represented by D 3 -D 0 3 =1.11x10 12 . exp(-445870/RT).t and D 3 -D 0 3 =2.55x10 9 .exp(-319240/RT).t for specimens with overall oxygen-to-metal ratios 1.97 and 2.0, respectively (activation energies expressed in J/mol). An example for the influence of the oxygen-to-metal ratio on the grain growth in mixed oxide fuel during operation in a fast reactor is also given. (orig.)

  6. Determination of thermodynamic properties and stability limit from fluorite phase of uranium and lanthanide mixed oxides, using galvanic cells with solid electrolytes

    International Nuclear Information System (INIS)

    Santiago, T.N.

    1980-10-01

    A method for thermodynamic properties determination for oxygen solubility in oxide systems at temperature interval 973 ≤ T [K] ≤ 1773 is described. A galvanic cell using as solid electrolytes zircon dioxide doped with 15% of calcium oxide is presented. This method was used for determining the phase change, temperature dependent, of uranium-lanthanides-oxygen Ln U O 4 stoichiometric system. (C.G.C.)

  7. Advanced fuel system technology for utilizing broadened property aircraft fuels

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    Possible changes in fuel properties are identified based on current trends and projections. The effect of those changes with respect to the aircraft fuel system are examined and some technological approaches to utilizing those fuels are described.

  8. Irradiation creep of the mixed oxide UPuO2

    International Nuclear Information System (INIS)

    Combette, Patrick; Milet, Claude

    1976-01-01

    The irradiation creep under compression of the mixed oxide UO 2 -PuO 2 was studied up to fission yields of 6x10 13 fcm -3 s -1 , under stresses -2 , in the temperature range 700-900 deg C. The creep rate is proportional to the applied stress and fission yield, athermal in the studied temperature range and non-dependent of burnup (up to 30000MWjt -1 ). In a sample under compression, swelling is observed due to the formation of fission products during the irradiation and the swelling rate is of the same order that in a cladded fuel element [fr

  9. Processing of Mixed Oxide Superconductors

    Science.gov (United States)

    1990-07-01

    rapid changes world wide a major research centre on high Tc superconductors was awarded to Cambridge which involved moving the work and people to a...reports and paper is in the appendices. Separation Ceramic superconductors tend to be mixtures of phases, especially when first discovered. It would...properties of the superconducting state will in principle allow superconducting material to be levitated from the non superconductor and several designs

  10. Transient redistribution of intragranular fission gas in irradiated mixed oxide

    International Nuclear Information System (INIS)

    Hinman, C.A.; Randklev, E.H.

    1981-01-01

    Safety analyses for an LMFBR require a knowledge of the fuel and fission gas behavior under transient conditions. Analyses of microstructural data derived from transiently heated, irradiated, mixed oxide fuel specimens have allowed the calculation of the degree of nonequilibrium of intragranular bubbles formed during the transient. It is hypothesized that the observed over-pressurization of the intragranular bubbles mechanically loads the fuel within the grain, leading to a stress gradient derived force upon near-grain-surface bubbles, driving them preferentially to the grain boundaries. Using existing models for forced diffusion it can be estimated that the stress derived forces on bubbles are within the same magnitude, and possibly greater, than the forces derived from the thermal gradient

  11. Plutonium-uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

    Science.gov (United States)

    Degueldre, C.; Martin, M.; Kuri, G.; Grolimund, D.; Borca, C.

    2011-09-01

    Plutonium-uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg -1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

  12. Bio energy: Bio fuel - Properties and Production

    International Nuclear Information System (INIS)

    Wilhelmsen, Gunnar; Martinsen, Arnold Kyrre; Sandberg, Eiliv; Fladset, Per Olav; Kjerschow, Einar; Teslo, Einar

    2001-01-01

    This is Chapter 3 of the book ''Bio energy - Environment, technique and market''. Its main sections are: (1) Definitions and properties, (2) Bio fuel from the forest, (3) Processed bio fuel - briquettes, pellets and powder, (4) Bio fuel from agriculture, (5) Bio fuel from agro industry, (6) Bio fuel from lakes and sea, (7) Bio fuel from aquaculture, (8) Bio fuel from wastes and (9) Hydrogen as a fuel. The exposition largely describes the conditions in Norway. The chapter on energy from the forest includes products from the timber and sawmill industry, the pulp and paper industry, furniture factories etc. Among agricultural sources are straw, energy forests, vegetable oil, bio ethanol, manure

  13. Using NJOY99 and MCNP4B2 to Estimate the Radiation Damage Displacements per Atom per Second in Steel Within the Boiling Water Reactor Core Shroud and Vessel Wall from Reactor-Grade Mixed-Oxide/Uranium Oxide Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

    International Nuclear Information System (INIS)

    Vickers, Lisa

    2003-01-01

    The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor

  14. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

    Hunter, C.W.; Johnson, G.D.

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  15. Power-to-melt evaluation of fresh mixed-oxide fast reactor fuel. Technical improvements of the post-irradiation-experiment and the evaluation of the results for the power-to-melt test PTM-2 in 'JOYO'

    International Nuclear Information System (INIS)

    Yamamoto, Kazuya; Kushida, Naoya; Koizumi, Atsuhiro

    1999-11-01

    The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor 'JOYO'. All of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. In this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at ∼24 mm below the peak power elevation in a test fuel pin. It is possible that this arose from secondary fuel-melting. Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well. (author)

  16. Characterization of manganese-gallium mixed oxide powders

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez Escribano, V.; Fernandez Lopez, E.; Sanchez Huidobro, P. [Universidad de Salamanca, Dept. de Quimica Inorganica (Spain); Panizza, M.; Resini, C.; Busca, G. [UNiversita di Genova, Dipt. di Ingegneria Chimica e di Processo, Genova (Italy); Resini, C. [Istituto Nazionale di Fisica della Materia, INFM (Spain); Gallardo- Amores, J.M. [Universidad Complutense, Dept. de Quimica Inorganica, Lab. Complutense de Altas Presiones, Madrid (Spain)

    2003-12-01

    Mn-Ga mixed oxides have been prepared by coprecipitation of the corresponding oxo-hydroxides as powders and have been characterized in relation to their structural and optical properties. The materials have been characterized by XRD, TG-DTA, skeletal IR and UV-visible-NIR spectroscopies. Large solubility of Mn in the diaspore type {alpha}-GaOOH oxo-hydroxide has been found. The spinel related structures of hausmannite Mn{sub 3}O{sub 4} and of {beta}-gallia present large reciprocal solubilities at least in a metastable form. At high temperature also bixbyite-type {alpha}-Mn{sub O3} solid solutions containing up to 20% at. Ga have been observed. (authors)

  17. Characterization of manganese?gallium mixed oxide powders

    Science.gov (United States)

    Sánchez Escribano, Vicente; Fernández López, Enrique; Sánchez Huidobro, Paula; Panizza, Marta; Resini, Carlo; Gallardo-Amores, José M.; Busca, Guido

    2003-11-01

    MnGa mixed oxides have been prepared by coprecipitation of the corresponding oxo-hydroxides as powders and have been characterized in relation to their structural and optical properties. The materials have been characterized by XRD, TG-DTA, skeletal IR and UV-visible-NIR spectroscopies. Large solubility of Mn in the diaspore type α-GaOOH oxo-hydroxide has been found. The spinel related structures of hausmannite Mn 3O 4 and of β-gallia present large reciprocal solubilities at least in a metastable form. At high temperature also bixbyite-type α-Mn 2O 3 solid solutions containing up to 20% at. Ga have been observed.

  18. Effect of metal ratio and calcination temperature of chromium based mixed oxides catalyst on FAME density from palm fatty acid distillate

    Science.gov (United States)

    Wan, Z.; Fatimah, S.; Shahar, S.; Noor, A. C.

    2017-09-01

    Mixed oxides chromium based catalysts were synthesized via sol-gel method for the esterification of palm fatty acid distillate (PFAD) to produce fatty acid methyl ester (FAME). The reactions were conducted in a batch reactor at reaction temperature of 160 °C for 4 h and methanol to PFAD molar ratio of 3:1. The effects of catalyst preparation conditions which are the mixed metal ratio and calcination temperature were studied. The various metal ratio of Cr:Mn (1:0, 0:1, 1:1, 1:2 and 2:1) and Cr:Ti (0:1, 1:1, 1:2 and 2:1) resulted in FAME density ranges from 1.041 g/cm3 to 0.853 g/cm3 and 1.107 g/cm3 to 0.836 g/cm3, respectively. The best condition catalyst was found to be Cr:Ti metal ratio of 1:2 and Cr:Mn metal ratio of 1:1. The calcination temperature of the mixed oxides between 300 °C to 700°C shows effect on the FAME density obtained in the reaction. The calcination at 500°C gave the lowest FAME density of 0.836 g/cm3 and 0.853 g/cm3 for Cr:Ti and Cr:Mn mixed oxides, respectively. The density of FAME is within the value range of the biodiesel fuel property. Thus, mixed oxides of Cr-Ti and Cr-Mn have good potentials as heterogeneous catalyst for FAME synthesis from high acid value oils such as PFAD.

  19. Yttrium bismuth titanate pyrochlore mixed oxides for photocatalytic hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Merka, Oliver

    2012-10-18

    In this work, the sol-gel synthesis of new non-stoichiometric pyrochlore titanates and their application in photocatalytic hydrogen production is reported. Visible light response is achieved by introducing bismuth on the A site or by doping the B site by transition metal cations featuring partially filled d orbitals. This work clearly focusses on atomic scale structural changes induced by the systematical introduction of non-stoichiometry in pyrochlore mixed oxides and the resulting influence on the activity in photocatalytic hydrogen production. The materials were characterized in detail regarding their optical properties and their atomic structure. The pyrochlore structure tolerates tremendous stoichiometry variations. The non-stoichiometry in A{sub 2}O{sub 3} rich compositions is compensated by distortions in the cationic sub-lattice for the smaller Y{sup 3+} cation and by evolution of a secondary phase for the larger Bi{sup 3+} cation on the A site. For TiO{sub 2} rich compositions, the non-stoichiometry leads to a special vacancy formation in the A and optionally O' sites. It is shown that pyrochlore mixed oxides in the yttrium bismuth titanate system represent very active and promising materials for photocatalytic hydrogen production, if precisely and carefully tuned. Whereas Y{sub 2}Ti{sub 2}O{sub 7} yields stable hydrogen production rates over time, the bismuth richer compounds of YBiTi{sub 2}O{sub 7} and Bi{sub 2}Ti{sub 2}O{sub 7} are found to be not stable under irradiation. This drawback is overcome by applying a special co-catalyst system consisting of a precious metal core and a Cr{sub 2}O{sub 3} shell on the photocatalysts.

  20. Thermal expansion studies on uranium-neodymium mixed oxide solid solutions

    International Nuclear Information System (INIS)

    Panneerselvam, G.; Venkata Krishnan, R.; Antony, M.P.; Nagarajan, K.

    2012-01-01

    Uranium-Neodymium mixed oxides solid solutions (U 1-y Nd y ) O 2 (y=0.2-0.95) were prepared by combustion synthesis using citric acid as fuel. Structural characterization and computation of lattice parameter was carried out from room temperature X-ray diffraction measurements. Single-phase fluorite structure was observed up to y=0.80. For solid solutions with y>0.80 additional Nd 2 O 3 lines were visible

  1. Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors

    International Nuclear Information System (INIS)

    Gonçalves, Letícia C.; Maiorino, José R.

    2017-01-01

    A comparative study of the mass balance and production costs of uranium oxide fuels was carried out for an AP1000 reactor and thorium-uranium mixed oxide in a reactor proposal using thorium called AP-Th1000. Assuming the input mass values for a fuel load the average enrichment for both reactors as well as their feed mass was determined. With these parameters, the costs were calculated in each fuel preparation process, assuming the prices provided by the World Nuclear Association. The total fuel costs for the two reactors were quantitatively compared with 18-month open cycle. Considering enrichment of 20% for the open cycle of mixed U-Th oxide fuel, the total uranium consumption of this option was 50% higher and the cost due to the enrichment was 70% higher. The results show that the use of U-Th mixed oxide fuels can be advantageous considering sustainability issues. In this case other parameters and conditions should be investigated, especially those related to fuel recycling, spent fuel storage and reduction of the amount of transuranic radioactive waste

  2. Characterization and Electrocatalytic Properties of Titanium-Based Ru0.3Co0.7−xCex Mixed Oxide Electrodes for Oxygen Evolution in Alkaline Solution

    Directory of Open Access Journals (Sweden)

    Hongjun Wu

    2011-01-01

    Full Text Available Ti-supported RuO2-Co3O4-CeO2 (Ru0.3Co0.7−xCex oxide, 0≤x≤0.7 electrodes were prepared by sol-gel process. The phase structure, surface morphology, and microstructure of the oxide layer were characterized by X-ray diffraction (XRD and scanning electron microscopy (SEM. Electrocatalytic activity and oxygen evolution reaction (OER kinetics on these electrodes in 1.0 mol⋅dm−3 KOH solution were studied by recording open-circuit potential, cyclic voltammetry, and polarisation curves. The results showed that the appropriate content of CeO2 could reduce the grain size and increase active surface area. The electrocatalytic activity shows a strong dependence on the CeO2 content in the film. Catalytic performance of mixed oxide electrodes with 40 mol % CeO2 was the best, with the greatest voltammetric charge, 86.23 mC⋅cm−2, and the smallest apparent activation energy for OER at 0.60 V was 22.76 kJ⋅mol−1.

  3. Thermal properties of heterogeneous fuels

    International Nuclear Information System (INIS)

    Staicu, D.; Beauvy, M.

    1998-01-01

    Fresh or irradiated nuclear fuels are composites or solid solutions more or less heterogeneous, and their thermal conductivities are strongly dependent on the microstructure. The effective thermal conductivities of these heterogeneous solids must be determined for the modelling of the behaviour under irradiation. Different methods (analytical or numerical) published in the literature can be used for the calculation of this effective thermal conductivity. They are analysed and discussed, but finally only few of them are really useful because the assumptions selected are often not compatible with the complex microstructures observed in the fuels. Numerical calculations of the effective thermal conductivity of various fuels based on the microstructure information provided in our laboratory by optical microscopy or electron micro-probe analysis images, have been done for the validation of these methods. The conditions necessary for accurate results on effective thermal conductivity through these numerical calculations are discussed. (author)

  4. Thermodynamic properties of the DUPIC fuel and its performance

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kwang Heon; Kim, Hee Moon [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-07-01

    This study describes thermodynamic properties of DUPIC fuel and performance. In initial state, DUPIC fuel which contains fissile materials is different from general nuclear fuel. So this study analyzed oxygen potential, thermal conductivity and specific heat of the DUPIC fuel.

  5. Study of the dissolution of (U,P)O2 mixed oxides with a high plutonium content

    International Nuclear Information System (INIS)

    Fournier, S.

    2001-01-01

    Plutonium from nuclear reactors is partially integrated in the fuel cycle as Mixed OXide (U,Pu)O 2 (MOX). Their dissolution in nitric acid is needed to reprocess them in present reprocessing plants. The main difficulty of this study is that dissolution is a phenomenon depending on solution characteristics as well as the structural properties of the pellets, which depend themselves on the material fabrication process. After showing kinetic and thermodynamic dissolution data of mixed oxides in nitric media, an inventory of the parameters which affect the dissolution process has been made. A separable variable concept was introduced in order to describe the process by studying separately the role of chemical parameters of the solution and geometric parameters of the material. The first part of the study estimates the effect of nitric solution chemical parameters (concentrations, acidity, temperature) on the dissolution and underlines the role of the oxide surface protonation step. The second part of this work deals with the study of surface area evolution for materials with controlled plutonium rich heterogeneities. Experimental results show that the pellet surface undergoes erosion and is progressively weakened by the formation of fault lines in the bulk of the material followed by the dispersion of sub-millimeter fragments in the solution. An heterogeneous kinetic model derived from study of solid-gas interface systems has been applied to fuel pellets dissolution, allowing a mechanism to be proposed, based on surface dissolution of the oxide as well as fault creation in the volume. The dissolution kinetics are therefore dependant on the microstructure and mechanical strength and cohesion of the pellets. (author)

  6. Preparation of mixed oxides (Th,U)O2: an evaluation of different techniques

    International Nuclear Information System (INIS)

    Ayoub, Jamil Mahmoud Said

    1999-01-01

    An evaluation of different ways of obtaining (Th-U)O 2 mixed oxides is described. Coprecipitation, mechanical mixing of uranium and thorium powders, and the sol-gel technique were studied in order to get a large spectrum of knowledge of the process performance. The use of ultrasonic waves for the homogenization of the hydroxide mixture and microwave heating for powder drying was also investigated. Sol-gel showed the best results regarding the specific area for obtained samples. Oxide drying by microwave is an effective method to get mixed oxides for fuel fabrication. Neither coprecipitation nor mechanical mixing of the thorium and uranium oxide powders is suitable for the purpose. The obtained data are less than 70% than those achieved when sol-gel process is performed. Electronic microscopy, X-ray fluorescence and diffraction, thermogravimetry, specific gravidity and specific area determination (BET) used for sample characterization were convenient and accomplished good results. (author)

  7. Thermophysical properties of reactor fuels

    International Nuclear Information System (INIS)

    Leibowitz, L.

    1981-01-01

    A review is presented of the literature on the enthalpy of uranium, thorium, and plutonium oxide and an approach is described for calculating the vapor pressure and gaseous composition of reactor fuel. In these calculations, thermodynamic functions of gas phase molecular species (obtained from matrix-isolation spectroscopy) are employed in conjunction with condensed phase therodynamics. A summary is presented of the status of this work

  8. Storage capacity and oxygen mobility in mixed oxides from transition metals promoted by cerium

    Energy Technology Data Exchange (ETDEWEB)

    Perdomo, Camilo [Estado Sólido y Catálisis Ambiental (ESCA), Departamento de Química, Facultad de Ciencias, Universidad Nacional de Colombia, Carrera 30 45-03, Bogotá (Colombia); Pérez, Alejandro [Grupo de Investigación Fitoquímica (GIFUJ), Departamento de Química, Facultad de Ciencias, Pontificia Universidad Javeriana, Carrera 7 No. 43-82, Bogotá D.C (Colombia); Molina, Rafael [Estado Sólido y Catálisis Ambiental (ESCA), Departamento de Química, Facultad de Ciencias, Universidad Nacional de Colombia, Carrera 30 45-03, Bogotá (Colombia); Moreno, Sonia, E-mail: smorenog@unal.edu.co [Estado Sólido y Catálisis Ambiental (ESCA), Departamento de Química, Facultad de Ciencias, Universidad Nacional de Colombia, Carrera 30 45-03, Bogotá (Colombia)

    2016-10-15

    Highlights: • Ce addition to the catalysts improves the availability of oxygen in the materials. • Mixed oxide with Co and Cu exhibits the best oxygen transport properties. • Co presence improves O{sub 2} mobility in the catalysts. • The presence of Cu in the solids improves redox properties. - Abstract: The oxygen mobility and storage capacity of Ce-Co/Cu-MgAl or Ce–MgAl mixed oxides, obtained by hydrotalcite precursors, were evaluated using Toluene-temperature-programmed-reaction, {sup 18}O{sub 2} isotopic exchange and O{sub 2}-H{sub 2} titration. The presence of oxygen vacancies-related species was evaluated by means of Electron Paramagnetic Resonance. A correlation was found between the studied properties and the catalytic activity of the oxides in total oxidation processes. It was evidenced that catalytic activity depends on two related processes: the facility with which the solid can be reduced and its ability to regenerate itself in the presence of molecular oxygen in the gas phase. These processes are enhanced by Cu-Co cooperative effect in the mixed oxides. Additionally, the incorporation of Ce in the Co-Cu catalysts improved their oxygen transport properties.

  9. Storage capacity and oxygen mobility in mixed oxides from transition metals promoted by cerium

    International Nuclear Information System (INIS)

    Perdomo, Camilo; Pérez, Alejandro; Molina, Rafael; Moreno, Sonia

    2016-01-01

    Highlights: • Ce addition to the catalysts improves the availability of oxygen in the materials. • Mixed oxide with Co and Cu exhibits the best oxygen transport properties. • Co presence improves O 2 mobility in the catalysts. • The presence of Cu in the solids improves redox properties. - Abstract: The oxygen mobility and storage capacity of Ce-Co/Cu-MgAl or Ce–MgAl mixed oxides, obtained by hydrotalcite precursors, were evaluated using Toluene-temperature-programmed-reaction, 18 O 2 isotopic exchange and O 2 -H 2 titration. The presence of oxygen vacancies-related species was evaluated by means of Electron Paramagnetic Resonance. A correlation was found between the studied properties and the catalytic activity of the oxides in total oxidation processes. It was evidenced that catalytic activity depends on two related processes: the facility with which the solid can be reduced and its ability to regenerate itself in the presence of molecular oxygen in the gas phase. These processes are enhanced by Cu-Co cooperative effect in the mixed oxides. Additionally, the incorporation of Ce in the Co-Cu catalysts improved their oxygen transport properties.

  10. Methane combustion over lanthanum-based perovskite mixed oxides

    Energy Technology Data Exchange (ETDEWEB)

    Arandiyan, Hamidreza [New South Wales Univ., Sydney (Australia). School of Chemical Engineering

    2015-11-01

    This book presents current research into the catalytic combustion of methane using perovskite-type oxides (ABO{sub 3}). Catalytic combustion has been developed as a method of promoting efficient combustion with minimum pollutant formation as compared to conventional catalytic combustion. Recent theoretical and experimental studies have recommended that noble metals supported on (ABO{sub 3}) with well-ordered porous networks show promising redox properties. Three-dimensionally ordered macroporous (3DOM) materials with interpenetrated and regular mesoporous systems have recently triggered enormous research activity due to their high surface areas, large pore volumes, uniform pore sizes, low cost, environmental benignity, and good chemical stability. These are all highly relevant in terms of the utilization of natural gas in light of recent catalytic innovations and technological advances. The book is of interest to all researchers active in utilization of natural gas with novel catalysts. The research covered comes from the most important industries and research centers in the field. The book serves not only as a text for researcher into catalytic combustion of methane, 3DOM perovskite mixed oxide, but also explores the field of green technologies by experts in academia and industry. This book will appeal to those interested in research on the environmental impact of combustion, materials and catalysis.

  11. Mixed oxide thermal behaviour at BOL: COMETHE III-J models and impact on power-to-melt

    International Nuclear Information System (INIS)

    Vliet, J. van

    1979-01-01

    The mixed oxide thermal behaviour at beginning of life is very important because it can impose a limitation to the fuel pin peak power, and therefore to the reactor thermal output. The relevant physical processes leading to fuel restructuring are modelled in COMETHE III-J in a kinetic way. This ensures that the temperature and power history are properly taken into account. These models are described and their impact on the calculated power to melt early in life is analysed. (author)

  12. Comparison of physical chemical properties of powders and respirable aerosols of industrial mixed uranium and plutonium oxide fuels

    International Nuclear Information System (INIS)

    Eidson, A.F.

    1982-01-01

    Studies were performed to characterize physical and chemical properties which may be important in determining the metabolism of accidentally released, inhaled aerosols of industrial mixed uranium and plutonium oxide fuels and to compare the properties of bulk powders and the respirable fraction they include. X-ray diffraction measurements showed that analysis of mixed-oxide powders from four process steps served to characterize their respirable fractions. IR spectroscopy was useful as a method to detect organic binders that were not observed by X-ray diffraction methods. Both X-ray diffraction and IR spectroscopy methods can be used in combination to identify the sources of a complex aerosol that might be released from more than one fabrication step. Isotopic distributions in powders and aerosols showed that information important for radiation dose to tissue calculations or Pu lung burden estimates can be obtained by analysis of powders. (U.K.)

  13. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    Fink, J.K.

    1984-01-01

    This paper identifies the fuel properties for which more data are needed for fast-reactor safety analysis. In addition, a brief review is given of current research on the vapor pressure over liquid UO 2 and (U,PU)O/sub 2-x/, the solid-solid phase transition in actinide oxides, and the thermal conductivity of molten urania

  14. Evaluation of thermal physical properties for fast reactor fuels. Melting point and thermal conductivities

    International Nuclear Information System (INIS)

    Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya

    2006-10-01

    Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (author)

  15. Effect of cooling rate on achieving thermodynamic equilibrium in uranium-plutonium mixed oxides

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C.; Robisson, Anne-Charlotte; Hodaj, Fiqiri

    2016-02-01

    In situ X-ray diffraction was used to study the structural changes occurring in uranium-plutonium mixed oxides U1-yPuyO2-x with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H2 atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s-1 and 0.005 K s-1, respectively. The promptly cooled samples evidenced a phase separation whereas samples cooled slowly did not due to their complete oxidation in contact with the atmosphere during cooling. Besides the composition of the annealing gas mixture, the cooling rate plays a major role on the control of the Oxygen/Metal ratio (O/M) and then on the crystallographic properties of the U1-yPuyO2-x uranium-plutonium mixed oxides.

  16. Standardization of thermal and epithermal INAA methods for simultaneous determination of U and Th in mixed oxide samples

    International Nuclear Information System (INIS)

    Acharya, R.; Pujari, P.K.; Chandra, Ruma

    2010-01-01

    Full text: Uranium and thorium are important fuel materials for nuclear power program. In recent years utilization of thoria based fuel has assumed significance due to higher energy requirements. Thorium based mixed oxide is the proposed fuel for Advanced Heavy Water Reactors (AHWR). In this respect, studies are carried out through preparation of natural U and Th mixed oxides by powder metallurgical route, wherein composition of U and Th is specific and requires strict control in terms their contents and homogeneity in the mixture. Stringent chemical quality control necessitates compositional characterization of the fuel material i.e. accurate and precise determination of U and Th. A suitable method which does not need any chemical dissolution and yields high precision results with minima sample handling is desirable. Instrumental neutron activation analysis (INAA) using reactor neutron is the technique of choice. In view of this, INAA methods namely thermal lNAA (TNAA) (utilizing whole reactor neutrons) and epithermal INAA (ENAA) (utilizing epicadmium neutrons) were standardized for the determination of U and Th in presence of each other in mixed oxide samples. In the present work pneumatic carrier facility (PCF) of Dhruva reactor and self-serve facility of CIRUS reactor were used for TNAA and ENAA respectively. Standards, synthetic samples and mixed oxide samples prepared in cellulose matrix, were irradiated for 1 minute at PCF of Dhruva reactor and for 1 hour at CIRUS reactor under cadmium cover (0.5 mm). Radioactive assay was carried out using 40% relative efficiency HPGe detector. Peak areas under the full energy peaks were evaluated by peak fit method using the PHAST software. Both activation and daughter products of U ( 239 U, 74.6 keV and 239 Np, 277 keV) and Th ( 233 Th, 86 keV and 233 Pa, 312 keV) were used for their concentration determination. The method was validated by analyzing synthetic mixed oxide samples (6-48%U-Th mixed oxide). The % deviations

  17. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    International Nuclear Information System (INIS)

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

  18. The effect of mixed oxidants and powdered activated carbon on the removal of natural organic matter.

    Science.gov (United States)

    Alvarez-Uriarte, Jon I; Iriarte-Velasco, Unai; Chimeno-Alanís, Noemí; González-Velasco, Juan R

    2010-09-15

    Present paper studies the influence of electrochemically generated mixed oxidants on the physicochemical properties of natural organic matter, and especially from the disinfection by-products formation point of view. The study was carried out in a full scale water treatment plant. Results indicate that mixed oxidants favor humic to non-humic conversion of natural organic matter. Primary treatment preferentially removes the more hydrophobic fraction. This converted the non-humic fraction in an important source of disinfection by-products with a 20% contribution to the final trihalomethane formation potential (THMFP(F)) of the finished water. Enhanced coagulation at 40 mg l(-1) of polyaluminium chloride with a moderate mixing intensity (80 rpm) and pH of 6.0 units doubled the removal efficiency of THMFP(F) achieved at full scale plant. However, gel permeation chromatography data revealed that low molecular weight fractions were still hardly removed. Addition of small amounts of powdered activated carbon, 50 mg l(-1), allowed reduction of coagulant dose by 50% whereas removal of THMFP(F) was maintained or even increased. In systems where mixed oxidants are used addition of powdered activated carbon allows complementary benefits by a further reduction in the THMFP(F) compared to the conventional only coagulation-flocculation-settling process. Copyright 2010 Elsevier B.V. All rights reserved.

  19. Iron-tellurium-selenium mixed oxide catalysts for the selective oxidation of propylene to acrolein

    International Nuclear Information System (INIS)

    Patel, B.M.; Price, G.L.

    1990-01-01

    This paper reports on iron-tellurium-selenium mixed oxide catalysts prepared by coprecipitation from aqueous solution investigated for the propylene to acrolein reaction in the temperature range 543-773 K. Infrared spectroscopy, electron dispersive X-ray analysis, X-ray diffraction, and isotopic tracer techniques have also been employed to characterize this catalytic system. Properties of the Fe-Te-Se mixed oxide catalysts have been compared with Fe-Te mixed oxides in an effort to deduce the functionality of Se. The selenium in the Fe-Te-Se-O catalyst has been found to be the hydrocarbon activating site. The activation energies for the acrolein and carbon dioxide formation are 71 and 54 kJ/mol, respectively. Reactions carried out with 18 O 2 have shown lattice oxygen to be primarily responsible for the formation of both acrolein and carbon dioxide. The initial and rate-determining step for acrolein formation is hydrogen abstraction as determined by an isotope effect associated with the C 3 D 6 reaction. No isotope effect is observed for carbon dioxide formation from C 3 D 6 suggesting that CO 2 is formed by parallel, not consecutive, oxidation of propylene

  20. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  1. Synthesis and characterization of composites of mixed oxides of iron ...

    Indian Academy of Sciences (India)

    Home; Journals; Bulletin of Materials Science; Volume 34; Issue 4. Synthesis and characterization of composites of mixed oxides of iron and neodymium in polymer matrix of aniline–formaldehyde. Sajdha H N Sheikh B L Kalsotra N Kumar S Kumar. Volume 34 Issue 4 July 2011 pp 843-851 ...

  2. Plutonium oxides and uranium and plutonium mixed oxides. Carbon determination

    International Nuclear Information System (INIS)

    Anon.

    Determination of carbon in plutonium oxides and uranium plutonium mixed oxides, suitable for a carbon content between 20 to 3000 ppm. The sample is roasted in oxygen at 1200 0 C, the carbon dioxide produced by combustion is neutralized by barium hydroxide generated automatically by coulometry [fr

  3. High-temperature thermophysical properties of nuclear fuels

    International Nuclear Information System (INIS)

    Hyland, G.J.; Ralph, Jeffrey

    1985-01-01

    This is a summary of discussions held at the 9th European Thermophysical Properties Conference. The first group discussed the following: the question of Bredig transition in UO 2 and related oxides, the thermal conductivity of molten UO 2 , the status of first principle calculations of the free energies of defect formation in UO 2 . A second group of topics discussed were: oxygen potentials over mixed oxide systems, the current status of vapor pressure measurements over liquid UO 2 made by use of laser heating techniques and their interpretation and preliminary results on electric and dielectric properties of solid UO 2 at high frequencies and low temperatures. The main interest was in the thermal conductivity of molten UO 2 and much of the discussion time was devoted to this. (U.K.)

  4. Remote mixed oxide fabrication facility development. Volume 2. State-of-the-art review of remote maintenance system technology

    International Nuclear Information System (INIS)

    Horgos, R.M.; Masch, M.L.

    1979-06-01

    This report provides a state-of-the-art review of remote systems technology, which includes manipulators, process connectors, vision systems and specialized process systems. A proposed mixed oxide fuel fabrication facility was reviewed and evaluated for identification of major remote maintenance and repair tasks. The technological areas were evaluated on the basis of their suitability or applicability for remote maintenance and repair of a proposed fully remote operating mixed oxide fuel fabrication facility. A technological base exists from which the design criteria for a reliable, remote operating facility can be established. Commercially available systems and components, along with those remote technologies now in development, will require modifications to adapt them to specific plant designs and requirements

  5. Property-process relationships in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Tikare, V.

    2015-01-01

    Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

  6. Fuel properties of loofah (Luffa cylindrica L.) biofuel blended with ...

    African Journals Online (AJOL)

    ajl6

    Fuel properties of loofah oil and its ethyl ester blended with diesel were experimentally determined. ... The escalating prices of petroleum fuels, the .... equation developed by Bamgboye and Hansen (2008) was used to ..... Renewable Energy.

  7. In vitro and in vivo measurements of the dissolution parameters of uranium and plutonium mixed oxides in biological environment

    International Nuclear Information System (INIS)

    Matton, S.

    1999-01-01

    During the mixed-oxide fuel fabrication process, inhalation is potentially the main route of internal contamination. The International Commission on Radiological Protection recommends experimental measurement of parameters such as size and dissolution rate for specific industrial compounds. First, we validated the use of PERALS (Photon Electron Rejecting Alpha Liquid Scintillation) for alpha measurement in biological samples which, in some cases, could improve detection limit. We characterised physical chemical properties in terms of size, specific area and activity of 3 different powders: MOX made according to either the MIMAS process, which showed heterogeneous chemical composition, or the SOLGEL, which showed homogeneous chemical composition and industrial PuO 2 . Their dissolution parameters, f r and s s , as defined in the simplest model proposed by ICRP 66 were measured in vivo, after inhalation in the rat, and in vitro. The statistical variation of these values were expressed as standard deviation. Moreover, in vitro studies demonstrated variation of the s s value depending on the duration of the incubation. We also developed methods to characterise interactions between UO 2 particles and phosphate ions which could be involved in the actinide toxicity. (author) [fr

  8. Amine–mixed oxide hybrid materials for carbon dioxide adsorption from CO2/H2 mixture

    Science.gov (United States)

    Ravi, Navin; Aishah Anuar, Siti; Yusuf, Nur Yusra Mt; Isahak, Wan Nor Roslam Wan; Shahbudin Masdar, Mohd

    2018-05-01

    Bio-hydrogen mainly contains hydrogen and high level of carbon dioxide (CO2). High concentration of CO2 lead to a limitation especially in fuel cell application. In this study, the amine-mixed oxide hybrid materials for CO2 separation from bio-hydrogen model (50% CO2:50% H2) have been studied. Fourier-transform infrared spectroscopy (FTIR) and x-ray diffraction (XRD) characterizations showed that the amine–mixed oxide hybrid materials successfully adsorbed CO2 physically with no chemical adsorption evidence. The dry gas of CO2/H2 mixture adsorbed physically on amine–CuO–MgO hybrid material. No carbonates were detected after several times of adsorption, which indicated the good recyclability of adsorbents. The adsorbent system of diethanolamine (DEA)/15% CuO–75% MgO showed the highest CO2 adsorption capacity of 21.2 wt% due to the presence of polar substance on MgO surface, which can adsorb CO2 at ambient condition. The alcohol group of DEA can enhance the CO2 solubility on the adsorbent surface. In the 20% CuO–50% MgO adsorbent system, DEA as amine type showed a high CO2 adsorption of 19.4 wt%. The 10% amine loading system showed that the DEA adsorption system provided high CO2 adsorption. The BET analysis confirmed that a high amine loading contributed to the decrease in CO2 adsorption due to the low surface area of the adsorbent system.

  9. Fuel property effects on Navy aircraft fuel systems

    Science.gov (United States)

    Moses, C. A.

    1984-01-01

    Problems of ensuring compatibility of Navy aircraft with fuels that may be different than the fuels for which the equipment was designed and qualified are discussed. To avoid expensive requalification of all the engines and airframe fuel systems, methodologies to qualify future fuels by using bench-scale and component testing are being sought. Fuel blends with increasing JP5-type aromatic concentration were seen to produce less volume swell than an equivalent aromatic concentration in the reference fuel. Futhermore, blends with naphthenes, decalin, tetralin, and naphthalenes do not deviate significantly from the correlation line of aromatic blends, Similar results are found with tensile strenth and elongation. Other elastomers, sealants, and adhesives are also being tested.

  10. Distillate Fuel Trends: International Supply Variations and Alternate Fuel Properties

    Science.gov (United States)

    2013-01-31

    fuel in NATO countries will have some amount of FAME present. There is some work being done on hydrocarbon alternatives but the regulatory structure ... synthesis or hydrotreatment – Requirements and test methods.” According to the specification, paraffinic diesel fuel does not meet the current requirements...or international specification for triglyceride based fuel oils (straight vegetable oil / raw vegetable oil). The same holds true for alcohol-based

  11. Phase relations, crystal structures and physical properties of nuclear fuels

    International Nuclear Information System (INIS)

    Tagawa, Hiroaki; Fujino, Takeo; Tateno, Jun

    1975-07-01

    Phase relations, crystal structures and physical properties of the compounds for nuclear fuels are presented, including melting point, thermal expansion, diffusion and magnetic and electric properties. Emphasis is on oxides, carbides and nitrides of thorium, uranium and plutonium. (auth.)

  12. Property measurements and inner state estimation of simulated fuel debris

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, S.; Kato, M.; Morimoto, K.; Washiya, T. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2014-07-01

    Fuel debris properties and inner state such as temperature profile were evaluated by using analysis of simulated fuel debris manufactured from UO{sub 2} and oxidized zircaloy. The center of the fuel debris was expected to be molten state soon after the melt down accident of LWRs because power density was very high. On the other hand, the surface of the fuel debris was cooled in the water. This large temperature gradient may cause inner stress and consequent cracks were expected. (author)

  13. CHARACTERIZATION OF MONOLITHIC FUEL FOIL PROPERTIES AND BOND STRENGTH

    International Nuclear Information System (INIS)

    D E Burkes; D D Keiser; D M Wachs; J S Larson; M D Chapple

    2007-01-01

    Understanding fuel foil mechanical properties, and fuel/cladding bond quality and strength in monolithic plates is an important area of investigation and quantification. Specifically, what constitutes an acceptable monolithic fuel--cladding bond, how are the properties of the bond measured and determined, and what is the impact of fabrication process or change in parameters on the level of bonding? Currently, non-bond areas are quantified employing ultrasonic determinations that are challenging to interpret and understand in terms of irradiation impact. Thus, determining mechanical properties of the fuel foil and what constitutes fuel/cladding non-bonds is essential to successful qualification of monolithic fuel plates. Capabilities and tests related to determination of these properties have been implemented at the INL and are discussed, along with preliminary results

  14. Energy properties of solid fossil fuels and solid biofuels

    International Nuclear Information System (INIS)

    Holubcik, Michal; Jandacka, Jozef; Kolkova, Zuzana

    2016-01-01

    The paper deals about the problematic of energy properties of solid biofuels in comparison with solid fossil fuels. Biofuels are alternative to fossil fuels and their properties are very similar. During the experiments were done in detail experiments to obtain various properties of spruce wood pellets and wheat straw pellets like biofuels in comparison with brown coal and black coal like fossil fuels. There were tested moisture content, volatile content, fixed carbon content, ash content, elementary analysis (C, H, N, S content) and ash fusion temperatures. The results show that biofuels have some advantages and also disadvantages in comparison with solid fossil fuels.

  15. Energy properties of solid fossil fuels and solid biofuels

    Energy Technology Data Exchange (ETDEWEB)

    Holubcik, Michal, E-mail: michal.holubcik@fstroj.uniza.sk; Jandacka, Jozef, E-mail: jozef.jandacka@fstroj.uniza.sk [University of Žilina, Faculty of Mechanical Engineering, Department of Power Engineering, Univerzitná 8215/1, 010 26 Žilina (Slovakia); Kolkova, Zuzana, E-mail: zuzana.kolkova@rc.uniza.sk [Research centre, University of Žilina, Univerzitna 8215/1, 010 26 Žilina (Slovakia)

    2016-06-30

    The paper deals about the problematic of energy properties of solid biofuels in comparison with solid fossil fuels. Biofuels are alternative to fossil fuels and their properties are very similar. During the experiments were done in detail experiments to obtain various properties of spruce wood pellets and wheat straw pellets like biofuels in comparison with brown coal and black coal like fossil fuels. There were tested moisture content, volatile content, fixed carbon content, ash content, elementary analysis (C, H, N, S content) and ash fusion temperatures. The results show that biofuels have some advantages and also disadvantages in comparison with solid fossil fuels.

  16. Modeling of Cd(II) sorption on mixed oxide

    International Nuclear Information System (INIS)

    Waseem, M.; Mustafa, S.; Naeem, A.; Shah, K.H.; Hussain, S.Y.; Safdar, M.

    2011-01-01

    Mixed oxide of iron and silicon (0.75 M Fe(OH)3:0.25 M SiO/sub 2/) was synthesized and characterized by various techniques like surface area analysis, point of zero charge (PZC), energy dispersive X-rays (EDX) spectroscopy, Thermogravimetric and differential thermal analysis (TG-DTA), Fourier transform infrared spectroscopy (FTIR) and X-rays diffraction (XRD) analysis. The uptake of Cd/sup 2+/ ions on mixed oxide increased with pH, temperature and metal ion concentration. Sorption data have been interpreted in terms of both Langmuir and Freundlich models. The Xm values at pH 7 are found to be almost twice as compared to pH 5. The values of both DH and DS were found to be positive indicating that the sorption process was endothermic and accompanied by the dehydration of Cd/sup 2+/. Further, the negative value of DG confirms the spontaneity of the reaction. The ion exchange mechanism was suggested to take place for each Cd/sup 2+/ ions at pH 5, whereas ion exchange was found coupled with non specific adsorption of metal cations at pH 7. (author)

  17. Effect of cooling rate on achieving thermodynamic equilibrium in uranium–plutonium mixed oxides

    Energy Technology Data Exchange (ETDEWEB)

    Vauchy, Romain, E-mail: romain.vauchy@cea.fr [CEA, DEN, DTEC, Marcoule, 30207, Bagnols-sur-Cèze (France); CEA, DEN, DEC, Cadarache, 13108, Saint-Paul-lez-Durance (France); Belin, Renaud C.; Robisson, Anne-Charlotte [CEA, DEN, DEC, Cadarache, 13108, Saint-Paul-lez-Durance (France); Hodaj, Fiqiri [Univ. Grenoble Alpes, SIMAP, F-38000, Grenoble (France); CNRS, Grenoble INP, SIMAP, F-38000, Grenoble (France)

    2016-02-15

    In situ X-ray diffraction was used to study the structural changes occurring in uranium–plutonium mixed oxides U{sub 1−y}Pu{sub y}O{sub 2−x} with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H{sub 2} atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s{sup −1} and 0.005 K s{sup −1}, respectively. The promptly cooled samples evidenced a phase separation whereas samples cooled slowly did not due to their complete oxidation in contact with the atmosphere during cooling. Besides the composition of the annealing gas mixture, the cooling rate plays a major role on the control of the Oxygen/Metal ratio (O/M) and then on the crystallographic properties of the U{sub 1−y}Pu{sub y}O{sub 2−x} uranium–plutonium mixed oxides.

  18. Separation and recovery of Cm from Cm-Pu mixed oxide samples containing Am impurity

    International Nuclear Information System (INIS)

    Hirokazu Hayashi; Hiromichi Hagiya; Mitsuo Akabori; Yasuji Morita; Kazuo Minato

    2013-01-01

    Curium was separated and recovered as an oxalate from a Cm-Pu mixed oxide which had been a 244 Cm oxide sample prepared more than 40 years ago and the ratio of 244 Cm to 240 Pu was estimated to 0.2:0.8. Radiochemical analyses of the solution prepared by dissolving the Cm-Pu mixed oxide in nitric acid revealed that the oxide contained about 1 at% of 243 Am impurity. To obtain high purity curium solution, plutonium and americium were removed from the solution by an anion exchange method and by chromatographic separation using tertiary pyridine resin embedded in silica beads with nitric acid/methanol mixed solution, respectively. Curium oxalate, a precursor compound of curium oxide, was prepared from the purified curium solution. 11.9 mg of Cm oxalate having some amounts of impurities, which are 243 Am (5.4 at%) and 240 Pu (0.3 at%) was obtained without Am removal procedure. Meanwhile, 12.0 mg of Cm oxalate (99.8 at% over actinides) was obtained with the procedure including Am removals. Both of the obtained Cm oxalate sample were supplied for the syntheses and measurements of the thermochemical properties of curium compounds. (author)

  19. Properties of light water reactor spent fuel cladding. Interim report

    International Nuclear Information System (INIS)

    Farwick, D.G.; Moen, R.A.

    1979-08-01

    The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

  20. Fuel properties to enable lifted-flame combustion

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, Eric [Ford Motor Company, Dearborn, MI (United States)

    2015-03-15

    The Fuel Properties to Enable Lifted-Flame Combustion project responded directly to solicitation DE-FOA-0000239 AOI 1A, Fuels and Lubricants for Advanced Combustion Regimes. This subtopic was intended to encompass clean and highly-efficient, liquid-fueled combustion engines to achieve extremely low engine-out nitrogen oxides (NOx) and particulate matter (PM) as a target and similar efficiency as state-of-the-art direct injection diesel engines. The intent of this project was to identify how fuel properties can be used to achieve controllable Leaner Lifted Flame Combustion (LLFC) with low NOx and PM emissions. Specifically, this project was expected to identify and test key fuel properties to enable LLFC and their compatibility with current fuel systems and to enhance combustion models to capture the effect of fuel properties on advanced combustion. Successful demonstration of LLFC may reduce the need for after treatment devices, thereby reducing costs and improving thermal efficiency. The project team consisted of key technical personnel from Ford Motor Company (FMC), the University of Wisconsin-Madison (UW), Sandia National Laboratories (SNL) and Lawrence Livermore National Laboratories (LLNL). Each partner had key roles in achieving project objectives. FMC investigated fuel properties relating to LLFC and sooting tendency. Together, FMC and UW developed and integrated 3D combustion models to capture fuel property combustion effects. FMC used these modeling results to develop a combustion system and define fuel properties to support a single-cylinder demonstration of fuel-enabled LLFC. UW investigated modeling the flame characteristics and emissions behavior of different fuels, including those with different cetane number and oxygen content. SNL led spray combustion experiments to quantify the effect of key fuel properties on combustion characteristics critical for LLFC, as well as single cylinder optical engine experiments to improve fundamental

  1. Novel Montmorillonite/TiO2/MnAl-Mixed Oxide Composites Prepared from Inverse Microemulsions as Combustion Catalysts

    Directory of Open Access Journals (Sweden)

    Bogna D. Napruszewska

    2017-11-01

    Full Text Available A novel design of combustion catalysts is proposed, in which clay/TiO2/MnAl-mixed oxide composites are formed by intermixing exfoliated organo-montmorillonite with oxide precursors (hydrotalcite-like in the case of Mn-Al oxide obtained by an inverse microemulsion method. In order to assess the catalysts’ thermal stability, two calcination temperatures were employed: 450 and 600 °C. The composites were characterized with XRF (X-ray fluorescence, XRD (X-ray diffraction, HR SEM (high resolution scanning electron microscopy, N2 adsorption/desorption at −196 °C, and H2 TPR (temperature programmed reduction. Profound differences in structural, textural and redox properties of the materials were observed, depending on the presence of the TiO2 component, the type of neutralization agent used in the titania nanoparticles preparation (NaOH or NH3 (aq, and the temperature of calcination. Catalytic tests of toluene combustion revealed that the clay/TiO2/MnAl-mixed oxide composites prepared with the use of ammonia showed excellent activity, the composites obtained from MnAl hydrotalcite nanoparticles trapped between the organoclay layers were less active, but displayed spectacular thermal stability, while the clay/TiO2/MnAl-mixed oxide materials obtained with the aid of NaOH were least active. The observed patterns of catalytic activity bear a direct relation to the materials’ composition and their structural, textural, and redox properties.

  2. Novel Montmorillonite/TiO₂/MnAl-Mixed Oxide Composites Prepared from Inverse Microemulsions as Combustion Catalysts.

    Science.gov (United States)

    Napruszewska, Bogna D; Michalik-Zym, Alicja; Rogowska, Melania; Bielańska, Elżbieta; Rojek, Wojciech; Gaweł, Adam; Wójcik-Bania, Monika; Bahranowski, Krzysztof; Serwicka, Ewa M

    2017-11-19

    A novel design of combustion catalysts is proposed, in which clay/TiO₂/MnAl-mixed oxide composites are formed by intermixing exfoliated organo-montmorillonite with oxide precursors (hydrotalcite-like in the case of Mn-Al oxide) obtained by an inverse microemulsion method. In order to assess the catalysts' thermal stability, two calcination temperatures were employed: 450 and 600 °C. The composites were characterized with XRF (X-ray fluorescence), XRD (X-ray diffraction), HR SEM (high resolution scanning electron microscopy, N₂ adsorption/desorption at -196 °C, and H₂ TPR (temperature programmed reduction). Profound differences in structural, textural and redox properties of the materials were observed, depending on the presence of the TiO₂ component, the type of neutralization agent used in the titania nanoparticles preparation (NaOH or NH₃ (aq)), and the temperature of calcination. Catalytic tests of toluene combustion revealed that the clay/TiO₂/MnAl-mixed oxide composites prepared with the use of ammonia showed excellent activity, the composites obtained from MnAl hydrotalcite nanoparticles trapped between the organoclay layers were less active, but displayed spectacular thermal stability, while the clay/TiO₂/MnAl-mixed oxide materials obtained with the aid of NaOH were least active. The observed patterns of catalytic activity bear a direct relation to the materials' composition and their structural, textural, and redox properties.

  3. Upgrading fuel properties of biomass by torrefaction

    Energy Technology Data Exchange (ETDEWEB)

    Lei Shang

    2012-12-15

    Torrefaction is a mild thermal (200 - 300 UC) treatment in an inert atmosphere, which is known to increase the energy density of biomass by evaporating water and a proportion of volatiles. In this work, the influence of torrefaction on the chemical and mechanical properties (grindability and hygroscopicity) of wood chips, wood pellets and wheat straw was investigated and compared. The mass loss during torrefaction was found to be a useful indicator for determining the degree of torrefaction. For all three biomass, higher torrefaction temperature or longer residence time resulted in higher mass loss, higher heating value, better grindability, and less moisture absorption. However, severe torrefaction conditions were found not necessary in order to save energy during grinding, since strain energy and grinding energy decreased tremendously in the first 5 - 25% anhydrous weight loss. By correlating the heating value and mass loss, it was found that wheat straw contained less heating value on mass basis than the other two fuels, but the fraction of energy retained in the torrefied sample as a function of mass loss was very similar for all three biomass. Gas products formed during torrefaction of three biomass were detected in situ by coupling mass spectrometer with a thermogravimetric analyzer (TGA). The main components were water, carbon monoxide, formic acid, formaldehyde, methanol, acetic acid, carbon dioxide, and methyl chloride. The cumulative releases of gas products from three biomass fuels at 300 UC for 1 h were compared, and water was found to be the dominant product during torrefaction. The degradation kinetics of wheat straw was studied in TGA by applying a two-step reaction in series model and taking the mass loss during the initial heating period into account. The model and parameters were proven to be able to predict the residual mass of wheat straw in a batch scale torrefaction reactor with different heating rates well. It means the mass yield of solids

  4. Low - temperature properties of rape seed oil biodiesel fuel and its blending with other diesel fuels

    International Nuclear Information System (INIS)

    Kampars, V.; Skujins, A.

    2004-01-01

    The properties of commercial bio diesel fuel depend upon the refining technique and the nature of the renewable lipids from which it is produced. The examined bio diesel fuel produced from rape seed oil by the Latvian SIA 'Delta Riga' has better low-temperature properties than many other bio diesels; but a considerably higher cloud point (-5,7 deg C), cold filter plugging point (-7 deg C) and pour point (-12 deg C) than the examined petrodiesel (grade C, LST EN 590:2000) from AB 'Mazeikiu nafta'. The low-temperature properties considerably improve if blending of these fuels is used. The blended fuels with bio diesel contents up to 90% have lower cold filter plugging points than petrodollar's. The estimated viscosity variations with temperature show that the blended fuels are Arrenius-type liquids, which lose this property near the cold filter plugging point. (authors)

  5. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

  6. Thorium utilization as a Pu-burner: proposal of Plutonium-Thorium Mixed Oxide (PT-MOX) Project

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    2000-01-01

    In this paper, a Pu-Th mixed oxide (PT-MOX) project is proposed for a thorium utilization and a plutonium burning. None of plutonium can be newly produced from PT-MOX fuel, and the plutonium mass of about 1 ton can be consumed with one reactor (total heavy metal assumed: 100 tons) for 1 year. In order to consume plutonium produced from usual Light Water Reactor, it should be better to operate one PT-MOX reactor for three to five Light Water Reactors. (author)

  7. A measurement of the density and compressibility of (U, Pu)-mixed oxide at 3432 kJ/kg (7356 K)

    International Nuclear Information System (INIS)

    Breitung, W.

    1987-05-01

    In a transient in-pile heating test the density of liquid (U, Pu)-mixed oxide at 3432±103 kJ/kg was measured to be 5027±132 kg/m 3 . The corresponding temperature is estimated to 7356±212 K. The isothermal compressibility of the mixed oxide was evaluated to 2.85 (±1.65) x 10 -4 /MPa at the same temperature. Based on these new data and previously existing measurements, new relations are proposed for the following properties of liquid UO 2 and (U, Pu)O 2 , as well: density - enthalpy, density - temperature, thermal expansion - temperature, and isothermal compressibility - temperature. (orig.) [de

  8. Wastes from the light water reactor fuel cycle

    International Nuclear Information System (INIS)

    Steindler, M.J.; Trevorrow, L.E.

    1976-01-01

    The LWR fuel cycle is represented, in the minimum detail necessary to indicate the origin of the wastes, as a system of operations that is typical of those proposed for various commercial fuel cycle ventures. The primary wastes (before any treatment) are described in terms of form, volume, radioactivity, chemical composition, weight, and combustibility (in anticipation of volume reduction treatments). Properties of the wastes expected from the operation of reactors, fuel reprocessing plants, and mixed oxide fuel fabrication plants are expressed in terms of their amounts per unit of nuclear energy produced

  9. Lithium containing MgAl mixed oxides obtained from sol-gel hydrotalcite for transesterification

    Directory of Open Access Journals (Sweden)

    Renata A. B. Lima-Corrêa

    Full Text Available Abstract The innumerous advantages of heterogeneous catalysts employed in biodiesel production have stimulated the search for a solid catalyst capable of replacing the industrially used homogeneous catalysts. This paper investigates the effect of the sol-gel method in the catalytic activity and stability of Li-MgAl mixed oxides prepared by the “in situ” lithium addition to a MgAl hydrotalcite. The analyses based on N2 physisorption, thermogravimetric analysis, X-ray diffractometry, scanning electron microscopy and temperature-programmed desorption of CO2 were carried out to elucidate the properties of the catalysts. Considerable differences in the physico-chemical properties of the catalysts were observed with the Li addition. Li reduced the surface area and increased the crystallite size of the oxides. Furthermore, Li-MgAl mixed oxides prepared by the calcination of the sol-gel MgAl hydrotalcites presented substantial morphological differences when compared to the same oxides obtained by heat treatment of hydrotalcites synthesized via the conventional co-precipitation route. Furthermore, Li increased the number and strength of the base sites which resulted in the increase of the oxide reactivities towards the transesterification reaction between methyl acetate and ethanol. The activity was dependent on the Li loading on the catalysts. The catalyst containing only 5 wt.% Li turned out to be highly active (( 85% conversion at 50°C, ethanol/methyl acetate molar ratio = 6/1, 4 wt.% of catalyst and 30 min of reaction. Stability tests showed that the Li-MgAl catalysts lose activity after 3 reuse cycles.

  10. Role of Cu-Mg-Al mixed oxide catalysts in lignin depolymerization in supercritical ethanol

    NARCIS (Netherlands)

    Huang, X.; Ceylanpinar, A.; Koranyi, T.I.; Boot, M.D.; Hensen, E.J.M.

    2015-01-01

    We investigate the role of Cu-Mg-Al mixed oxides in depolymerization of soda lignin in supercritical ethanol. A series of mixed oxides with varying Cu content and (Cu+Mg)/Al ratio were prepared. The optimum catalyst containing 20 wt% Cu and having a (Cu+Mg)/Al ratio of 4 yielded 36 wt% monomers

  11. A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Adams, Marvin L.

    2002-01-01

    We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

  12. Chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade mixed oxides [(U,Pu)O2

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    Mixed oxide, a mixture of uranium and plutonium oxides, is used as a nuclear-reactor fuel in the form of pellets. The plutonium content may be up to 10 wt %, and the diluent uranium may be of any U-235 enrichment. In order to be suitable for use as a nuclear fuel, the material must meet certain criteria for combined uranium and plutonium content, effective fissile content, and impurity content. Analytical procedures used to determine if mixed oxides comply with specifications are: uranium by controlled-potential coulometry; plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron (II); nitrogen by distillation spectrophotometry using Nessler reagent; carbon (total) by direct combustion-thermal-conductivity; total chlorine and fluorine by pyrohydrolysis; sulfur by distillation-spectrophotometry; moisture by the coulometric, electrolytic moisture analyzer; isotopic composition by mass spectrometry; rare earths by copper spark spectroscopy; trace impurities by carrier distillation spectroscopy; impurities by spark-source mass spectrography; total gas in reactor-grade mixed dioxide pellets; tungsten by dithiol-spectrophotometry; rare earth elements by spectroscopy; plutonium-238 isotopic abundance by alpha spectrometry; uranium and plutonium isotopic analysis by mass spectrometry; oxygen-to-metal atom ratio by gravimetry

  13. Advanced technique for computing fuel combustion properties in pulverized-fuel fired boilers

    Energy Technology Data Exchange (ETDEWEB)

    Kotler, V.R. (Vsesoyuznyi Teplotekhnicheskii Institut (Russian Federation))

    1992-03-01

    Reviews foreign technical reports on advanced techniques for computing fuel combustion properties in pulverized-fuel fired boilers and analyzes a technique developed by Combustion Engineering, Inc. (USA). Characteristics of 25 fuel types, including 19 grades of coal, are listed along with a diagram of an installation with a drop tube furnace. Characteristics include burn-out intensity curves obtained using thermogravimetric analysis for high-volatile bituminous, semi-bituminous and coking coal. The patented LFP-SKM mathematical model is used to model combustion of a particular fuel under given conditions. The model allows for fuel particle size, air surplus, load, flame height, and portion of air supplied as tertiary blast. Good agreement between computational and experimental data was observed. The method is employed in designing new boilers as well as converting operating boilers to alternative types of fuel. 3 refs.

  14. Analysis of the porosity distribution of mixed oxide pins; Analisis de distribucion de porosidad en barras combustibles de oxidos mixtos bajo irradiacion

    Energy Technology Data Exchange (ETDEWEB)

    Lieblich, M; Lopez, J

    1987-07-01

    In the frame of the Joint Irradiation Program IVO-FR2-Vg7 between the Centre of Nuclear Research of Karlsruhe (KfK), the irradiation of 30 mixed-oxide fuel rods in the FR2 experimental reactor was carried out. The pins were located in 10 single-walled NaK capsules. The behaviour of the fuel during its burnup was studied, mainly, the rest-porosity and cracking distribution in the pellet, partial densification, etc. In this work 3 pins from the capsule No. 165 were analyzed. The experimental results (pore and cracking profiles) were interpreted by the fuel rod code SATURN. (Author) 20 refs.

  15. Basic properties of fuel determining its behavior under irradiation

    International Nuclear Information System (INIS)

    Konovalov, I.I.

    2000-01-01

    The theoretical model describing a swelling of nuclear fuel at low irradiation temperatures is considered. The critical physical parameters of substances determining behavior of point defects, gas fission atoms, dislocation density, nucleation and growth of gas-contained pores are determined. The correlation between meanings of critical parameters and physical properties of substance is offered. The accounts of swelling of various dense fuels with reference to work in conditions of research reactors are given. (author)

  16. Assessment of microalgae biodiesel fuels using a fuel property estimation methodology

    Energy Technology Data Exchange (ETDEWEB)

    Torrens, Jonas Colen Ladeia; Vargas, Jose Viriato Coelho; Mariano, Andre Bellin [Center for Research and Development of Sustainable Energy. Universidade Federal do Parana, Curitiba, PR (Brazil)

    2010-07-01

    Recently, depleting supplies of petroleum and the concerns about global warming are drawing attention to alternative sources of energy. In this context, advanced biofuels, derived from non edible superior plants and microorganisms, are presented as promising options for the transportation sector. Biodiesel, which is the most prominent alternative fuel for compression ignition engines, have a large number as potential feedstock, such as plants (e.g., soybean, canola, palm) and microorganism (i.e., microalgae, yeast, fungi and bacterium). In order to determine their potential, most studies focus on the economic viability, but few discuss the technical viability of producing high quality fuels from such feedstock. Since the fuel properties depend on the composition of the parent oil, and considering the variability of the fatty acid profile found in these organisms, it is clear that the fuels derived may present undesirable properties, e.g., high viscosity, low cetane number, low oxidative stability and poor cold flow properties. Therefore, it is very important to develop ways of analysing the fuel quality prior to production, specially considering the high cost of producing and testing several varieties of plants and microorganisms. In this aim, this work presents the use of fuel properties estimation methods on the assessment of the density, viscosity, cetane number and cold filter plugging point of several microalgae derived biofuels, comparing then to more conventional biodiesel fuels. The information gathered with these methods helps on the selection of species and cultivation parameters, which have a high impact on the derived fuel quality, and have been successfully employed on the Center for Research and Development of Sustainable Energy. The results demonstrate that some species of microalgae have the potential to produce high quality biodiesel if cultivated with optimised conditions, associated with the possibility of obtaining valuable long chain

  17. Ultrasonic measurement of high burn-up fuel elastic properties

    International Nuclear Information System (INIS)

    Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

    2006-01-01

    The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

  18. Properties of zirconium carbide for nuclear fuel applications

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

    2013-10-15

    Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

  19. Some properties for modeling of fuel elements

    International Nuclear Information System (INIS)

    Nichols, F.A.

    1979-01-01

    Two areas key to the materials modeling of fuel element behavior are discussed. The relative importance of atomic diffusion vs. bubble migration is first surveyed and the interplay of bubble mobility and re-solution parameter is highlighted. It is concluded that biased bubble migration at higher temperatures is required to explain available gas-release data, especially during transients. At intermediate temperatures, random bubble migration is required to explain both gas-release rates and the observation of large (approx. 700A) intragranular bubbles following in-pile and post-irradiation transients. Different fuel models employ different values of re-solution parameter, both below and above an experimentally determined value. Bubble mobilities are deduced to approach theoretical, surface diffusion-controlled values during transients, but they may be somewhat less mobile during steady-state operation. Next, the present understanding of radiation-induced hardening and creep is discussed, highlighting the interplay of these two phenomena. An overall constitutive scheme is presented and predictions of failure limits are deduced therefrom employing instability analysis

  20. CERMET fuel behavior and properties in ADS reactors

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Staicu, D.; Somers, J.; Maschek, W.; Liu, P.; Chen, X.

    2008-01-01

    Within the EUROTRANS Integrated Project, Forschungszentrum Karlsruhe (FZK) and the Institute for Transuranium Elements (ITU) are joining their efforts to study the behavior of Mo-based CERMET non-uranium fuel for the ADS. Contributions include core safety calculations, and fuel property measurements and irradiation experiments. Safety studies for optimized EFIT core designs have concluded that, for the new low power cores of EFIT with a power class of ∼400 MWth and a fuel power density of ∼250 MW/m 3 , the CERMET-loaded cores behave favorably and the design limits of the fuels were not violated. Mo-based CERMET fuel pellets and pins loaded with Pu and Am were fabricated for irradiation programmes which will start by mid-2007 in PHENIX (France) and HFR-Petten (The Netherlands). The thermal diffusivity and specific heat of the CERMET fuels (loaded with Pu and Am) were the main properties measured, and the thermal conductivity was deduced. The results were used to prepare the safety report for the irradiation experiments

  1. Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

    Energy Technology Data Exchange (ETDEWEB)

    Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

    1997-12-01

    This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory.

  2. Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

    International Nuclear Information System (INIS)

    Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

    1997-12-01

    This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory

  3. Insight into the photocatalytic activity of ZnCr-CO3 LDH and derived mixed oxides

    Czech Academy of Sciences Publication Activity Database

    Paušová, Š.; Krysa, J.; Jirkovský, Jaromír; Forano, C.; Mailhot, G.; Prevot, V.

    2015-01-01

    Roč. 170, JUL 01 (2015), s. 25-33 ISSN 0926-3373 Institutional support: RVO:61388955 Keywords : photocatalysis * layered double hydroxides * mixed oxides Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 8.328, year: 2015

  4. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  5. Preparation of mixed oxides (Th,U)O{sub 2}: an evaluation of different techniques; Estudos de diferentes rotas de preparacao de oxidos binarios de torio e uranio

    Energy Technology Data Exchange (ETDEWEB)

    Ayoub, Jamil Mahmoud Said

    1999-07-01

    An evaluation of different ways of obtaining (Th-U)O{sub 2} mixed oxides is described. Coprecipitation, mechanical mixing of uranium and thorium powders, and the sol-gel technique were studied in order to get a large spectrum of knowledge of the process performance. The use of ultrasonic waves for the homogenization of the hydroxide mixture and microwave heating for powder drying was also investigated. Sol-gel showed the best results regarding the specific area for obtained samples. Oxide drying by microwave is an effective method to get mixed oxides for fuel fabrication. Neither coprecipitation nor mechanical mixing of the thorium and uranium oxide powders is suitable for the purpose. The obtained data are less than 70% than those achieved when sol-gel process is performed. Electronic microscopy, X-ray fluorescence and diffraction, thermogravimetry, specific gravidity and specific area determination (BET) used for sample characterization were convenient and accomplished good results. (author)

  6. Yttrium aluminum garnet (YAG) obtained by rare-earth mixed oxide (RE2O3)

    International Nuclear Information System (INIS)

    Castro, D.F.; Daguano, J.K.M.F.; Rodrigues Junior, D.; Suzuki, P.A.; Silva, O.M.M.

    2010-01-01

    In this work, the substitution of commercial Y 2 O 3 by a rare earth mixed oxide, RE 2 O 3 , to form Yttrium aluminum Garnet-Y 3 Al 5 O 12 , was investigated. Al 2 O 3 :Y 2 O 3 and Al 2 O 3 :RE 2 O 3 powder-mixtures, in a molar ratio of 60:40, were milled and subsequently cold uniaxially-pressed. Compacts were sintered at 1000, 1400 or 1600 deg C, for 120 minutes. RE 2 O 3 oxide was characterized by high-resolution synchrotron X-ray diffraction (HRXRD) and compared to Y 2 O 3 . X-ray diffraction pattern of the RE 2 O 3 indicates a true solid solution formation. Rietveld refinement of the sintered YAG and (RE)AG reveled a similar crystal structure to the YAGs obtained by the use of Al 2 O 3 -Y 2 O 3 or Al 2 O 3 -RE 2 O 3 respectively. Microstructural analysis of both, YAG or (RE)AG, revealed similar grain sizes of about 2.5 μm besides mechanical properties, with hardness of 400HV and fracture toughness of 3.8MPa.m1/2. It could be, thus, demonstrated that pure Y 2 O 3 can be substituted by the rare-earth solid solution, RE 2 O 3 , in the formation YAGs, presenting similar microstructural and mechanical properties. (author)

  7. Enhanced photocatalytic activity of titania-silica mixed oxide prepared via basic hydrolyzation

    International Nuclear Information System (INIS)

    Xie Chao; Xu Zili; Yang Qiujing; Xue Baoyong; Du Yaoguo; Zhang Jiahua

    2004-01-01

    Two different synthesis routes were applied to prepare TiO 2 -XSiO 2 (X denotes mol% of silica in titania-silica mixed oxides) with different silica concentrations by using ammonia water as hydrolysis catalyst. Through comparing the photocatalytic performance of two sets of mixed oxides, we found that the photocatalytic activity of mixed oxides prepared via the route which can promote homogeneity, was significantly enhanced as compared with that of counterparts prepared via the another route, and the highest photocatalytic activity obtained by adding about 9.1 mol% silica into titania was much higher than that of pure TiO 2 . The mixed oxides were investigated by means of XRD, thermal analysis, UV-vis, FT-IR and XPS. The characterization results suggest that, in comparison with pure TiO 2 , the mixed oxides exhibit smaller crystallite size and higher thermal stability which can elevate the temperature of anatase to rutile phase transformation due to the addition of silica. Furthermore, Broensted acidity, which is associated with the formation of Ti-O-Si hetero linkages where tetrahedrally coordinated silica is chemically mixed with the octahedral titania matrix, may be a very important contribution to the enhanced photocatalytic activity of titania-silica mixed oxides as well

  8. Study of fuel properties of rubber seed oil based biodiesel

    International Nuclear Information System (INIS)

    Ahmad, Junaid; Yusup, Suzana; Bokhari, Awais; Kamil, Ruzaimah Nik Mohammad

    2014-01-01

    Graphical abstract: - Highlights: • This article presents the comparative studies of the fuel properties of rubber seed oil based biodiesel. • The design expert has been adopted for the optimization of the process variables. • The FTIR, cold flow properties and oxidation stability are the findings of present study. • All the fuel properties met the standards such as ASTM D6751 and EN 14214. • Present study reveals that rubber seed oil as a non-edible source potentially contributes for esters production. - Abstract: The scarcity of the fossil fuel, environmental pollution and food crisis are the world’s major issues in current era. Biodiesel is an alternative to diesel fuel, environment friendly and biodegradable and is produced from either edible or non-edible oils. In this study, a non-edible rubber seed oil (RSO) with high free fatty acid (FFA) content of 45% were used for the production of biodiesel. The process comprises of two steps. The first step is the acid esterification to reduce the FFA value and the second step is the base transesterification. The response surface methodology (RSM) was used for parametric optimization of the two stage processes i.e. acid esterification and base transesterification. The yield of biodiesel was analyzed using gas chromatography. The FTIR (Fourier Transform Infra-Red) spectrum was also determined to confirm the conversion of fatty acid to methyl esters. The fuel properties were analyzed according to the ASTM D6751 and EN14214 and were compared with the previous finding of researchers. All analyzed properties fulfilled the biodiesel standard criteria

  9. Synthesis and Characterization of K-Ta Mixed Oxides for Hydrogen Generation in Photocatalysis

    Directory of Open Access Journals (Sweden)

    Beata Zielińska

    2012-01-01

    Full Text Available K-Ta mixed oxides photocatalysts have been prepared by impregnation followed by calcination. The influence of the reaction temperature (450°C–900°C on the phase formation, crystal morphology, and photocatalytic activity in hydrogen generation of the produced materials was investigated. The detailed analysis has revealed that all products exhibit high crystallinity and irregular structure. Moreover, two different crystal structures of potassium tantalates such as KTaO3 and K2Ta4O11 were obtained. It was also found that the sample composed of KTaO3 and traces of unreacted Ta2O5 (annealed at 600°C exhibits the highest activity in the reaction of photocatalytic hydrogen generation. The crystallographic phases, optical and vibronic properties were examined by X-ray diffraction (XRD and diffuse reflectance (DR UV-vis and resonance Raman spectroscopic methods, respectively. Morphology and chemical composition of the produced samples were studied using a high-resolution transmission electron microscope (HR-TEM and an energy dispersive X-ray spectrometer (EDX as its mode.

  10. Synthesis and characterization of Tin / Titanium mixed oxide nanoparticles doped with lanthanide for biomarking

    International Nuclear Information System (INIS)

    Paganini, Paula Pinheiro

    2012-01-01

    This work presents the synthesis, characterization and photo luminescent study of tin and titanium mixed oxide nanoparticles doped with europium, terbium and neodymium to be used with luminescent markers on biological systems. The syntheses were done by co-precipitation, protein sol-gel and Pechini methods and the nanoparticles were characterized by infrared spectroscopy, thermogravimetric analysis, scanning electron microscopy, X-ray diffraction and X-ray absorption spectroscopy. The photo luminescent properties studies were conducted for luminophores doped with europium, terbium and neodymium synthesized by coprecipitation method. For luminophore doped with europium it was possible to calculate the intensity parameters and quantum yield and it showed satisfactory results. In the case of biological system marking it was necessary the functionalization of these particles to allow them to bind to the biological part to be studied. So the nanoparticles were functionalized by microwave and Stöber methods and characterized by infrared spectroscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy and X-ray diffraction obtaining qualitative response of functionalization efficacy. The ninhydrin spectroscopic method was used for quantification of luminophores functionalization. The photo luminescent studies of functionalized particles demonstrate the potential applying of these luminophores as luminescent markers. (author)

  11. Micromechanical modelling of fuel viscoplastic behaviour

    International Nuclear Information System (INIS)

    Masson, R.; Blanc, V.; Gatt, J.M.; Julien, J.; Michel, B.; Largenton, R.

    2015-01-01

    To identify the effect of microstructural parameters on the viscoplastic behaviour of nuclear fuels, micromechanical (also called homogenisation) approaches are used. These approaches aim at deriving effective properties of heterogeneous material from the properties of their constituents. They stand on full-field computations of representative volume elements of microstructures as well as on mean-field semi-analytical models. For light water reactor fuels, these approaches have been applied to the modelling of the effect of two microstructural parameters: the porosity effects on the thermal creep of dioxide uranium fuels (transient conditions of irradiation) as well as the plutonium content effect on the viscoplastic behaviour (nominal conditions of irradiations) of mixed oxide fuels (MOX). (authors)

  12. Calculations of thermodynamic properties of PuO{sub 2} by the first-principles and lattice vibration

    Energy Technology Data Exchange (ETDEWEB)

    Minamoto, Satoshi [Energy and Industrial Systems Department, ITOCHU Techno-Solutions Corporation, Kasumigaseki 3-chome, Chiyoda-ku, Tokyo 100-6080 (Japan)], E-mail: satoshi.minamoto@ctc-g.co.jp; Kato, Masato [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan); Konashi, Kenji [Institute for Materials Research, Tohoku University, 2145-2 Narita-chou, Oarai-chou, Ibaraki 311-1313 (Japan); Kawazoe, Yoshiyuki [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai 980-8577 (Japan)

    2009-03-15

    Plutonium dioxide (PuO{sub 2}) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO{sub 2} at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO{sub 2} were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO{sub 2} were investigated up to 1500 K.

  13. Calculations of thermodynamic properties of PuO2 by the first-principles and lattice vibration

    International Nuclear Information System (INIS)

    Minamoto, Satoshi; Kato, Masato; Konashi, Kenji; Kawazoe, Yoshiyuki

    2009-01-01

    Plutonium dioxide (PuO 2 ) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO 2 at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO 2 were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO 2 were investigated up to 1500 K

  14. Automated x-ray spectrometer for mixed oxide pellets

    International Nuclear Information System (INIS)

    Lambert, M.C.; Goheen, M.W.; Urie, M.W.; Wynhoff, N.

    1979-01-01

    This paper discusses the development of an energy dispersive x-ray (EDX) spectrometer for the rapid, automated, close-coupled analysis of solid mixed plutonium--uranium oxide fuel pellets. Reasons are given for the system design, which is intended to give a relative precision (RSD) of +-0.3% in a total analysis time of three minutes. The principal problems in an EDX system are in maximizing the plutonium count rates

  15. Rational design of Mg-Al mixed oxide-supported bimetallic catalysts for dry reforming of methane

    Energy Technology Data Exchange (ETDEWEB)

    Tsyganok, Andrey I. [Centre for Catalysis Research and Innovation, Department of Chemistry, University of Ottawa, D' Iorio Hall, 10 Marie Curie Street, Ottawa, Ont. (Canada); Inaba, Mieko [Natural Gas Technology Development Team, Teikoku Oil Co., 9-23-30 Kitakarasuyama, Setagaya-ku, Tokyo 157-0061 (Japan); Tsunoda, Tatsuo; Uchida, Kunio; Suzuki, Kunio; Hayakawa, Takashi [Institute for Materials and Chemical Process, National Institute of Advanced Industrial Science and Technology (AIST), Tsukuba Central 5, 1-1-1 Higashi, Tsukuba 305-8565 (Japan); Takehira, Katsuomi [Department of Applied Chemistry, Faculty of Engineering, Hiroshima University, 1-4-1 Kagamiyama, Higashi-Hiroshima 739-8527 (Japan)

    2005-09-18

    A novel synthetic strategy for preparing bimetallic Ru-M (M=Cr, Fe, Co, Ni and Cu) catalysts, supported on Mg-Al mixed oxide, has been introduced. It was based on a 'memory effect', i.e. on the ability of Mg-Al mixed oxide to reconstruct a layered structure upon rehydration with an aqueous solution. By repeated calcinations-rehydration cycles, layered double hydroxide (LDH) precursors of catalysts containing two different metals were synthesized. Bimetallic catalysts were then generated (1) in situ from LDH under methane reforming reaction conditions and (2) from mixed metal oxides obtained by preliminary LDH calcination. Among all the LDH-derived catalysts, a Ru{sup 0.1%}-Ni{sup 5.0%}/MgAlO{sub x} sample revealed the highest activity and selectivity to syngas, a suitable durability and a low coking capacity. A promoting effect of ruthenium on catalytic function of supported nickel was demonstrated. Preliminary LDH calcination was shown to markedly affect the catalytic activity of the derived catalysts and especially their coking properties.

  16. Synthesis, characterization and photocatalytic activity of mixed oxides derived from ZnAlTi ternary layered double hydroxides

    Science.gov (United States)

    Sahu, R. K.; Mohanta, B. S.; Das, N. N.

    2013-09-01

    A new series of Ti4+ containing ZnAl-LDHs with varying Zn:Al:Ti (~3:1:0-3:0.5:0.5) ratio were prepared by coprecipitation of homogeneous solution metal salts and characterized by various physicochemical methods. Powder XRD revealed the formation of well crystallized LDH even at the highest Ti4+ content. On thermal treatment at 450 °C, the well crystallized LDH precursors yielded mixed oxides with BET surface area in the range 92-118 m2/g. UV-vis diffuse reflection spectroscopy (DRS) showed a marginal decrease of band gap energy for calcined ZnAlTi-LDHs in comparison to either ZnO or TiO2-P25. The TEM analyses of a representative sample (as-synthesized and calcined) indicated more or less uniform distribution of titanium species. The derived mixed oxides from titanium containing LDH precursors demonstrated better activity toward photodegradation of methylene blue and rhodamine B than those of a physical mixture of ZnO and TiO2. Moreover, the present work not only provided a first hand understanding about semiconductor properties of ZnAlTi-LDHs but also demonstrated their potential as photocatalysts for degradation of organic pollutants.

  17. Construction of database server system for fuel thermo-physical properties

    International Nuclear Information System (INIS)

    Park, Chang Je; Kang, Kwon Ho; Song, Kee Chan

    2003-12-01

    To perform the evaluation of various fuels in the nuclear reactors, not only the mechanical properties but also thermo-physical properties are required as one of most important inputs for fuel performance code system. The main objective of this study is to make a database system for fuel thermo-physical properties and a PC-based hardware system has been constructed for ease use for the public with visualization such as web-based server system. This report deals with the hardware and software which are used in the database server system for nuclear fuel thermo-physical properties. It is expected to be highly useful to obtain nuclear fuel data without such a difficulty through opening the database of fuel properties to the public and is also helpful to research of development of various fuel of nuclear industry. Furthermore, the proposed models of nuclear fuel thermo-physical properties will be enough utilized to the fuel performance code system

  18. Cermet fuel behaviour and properties in ADS reactors

    International Nuclear Information System (INIS)

    Haas, D.; Fernandez, A.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

    2007-01-01

    Within the EUROTRANS Integrated Project co- financed within the 6th Framework Programme of the European commission, the sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. Within the EUROTRANS Programme, the domain AFTRA is responsible to develop and provide the data basis for the fuels to be used in the European Facility for Industrial Transmutation (EFIT). The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmuter fuels are strongly linked with the core design and safety parameters, the fuel properties and the ease of fabrication and reprocessing. This study concerns the behaviour and properties of fuels with molybdenum as inert matrix. The status of the development work was presented at the last ICENES conference [1]. Since then, the design of the European Facility for Industrial Transmutation (EFIT) was developed and the transmutation capability, the burn-up behaviour, the reactivity swing and power peaking factors, and the safety performance were determined for different cores with inert matrix fuels like MgO and Mo. For the EFIT, the CERMET with the Mo matrix is recommended as the reference fuel and CERCER with the MgO matrix as a back-up solution. The thermal diffusivity and specific heat of the CERMET fuels (loaded with Pu and Am) were measured, and the thermal conductivity was deduced. The thermal conductivity of the CERMET fuels was also predicted using a model proposed in [1], with a microstructure corresponding to a random distribution of spheres, with overlapping. This model microstructure takes into account the negative effects arising from the possible formation of small agglomerates of inclusions in the CERMET fuels. The agreement between the theoretical and calculated values is relatively good (the error is between 0 and 15% of the value of the thermal conductivity

  19. Late effects following inhalation of mixed oxide (U,PuO2) mox aerosol in the rat

    International Nuclear Information System (INIS)

    Griffiths, N.; Van Der Meeren, A.; Fritsch, P.; Maximilien, R.

    2008-01-01

    Exposure to alpha-emitting particles is a potential long-term health risk to workers in nuclear fuel fabrication plants. Mixed Oxide (MOX: U,PuO 2 ) fuels containing low percentages of plutonium obtained from spent nuclear fuels are increasingly employed and in the case of accidental contamination by inhalation or wounds may result in the development of late-occurring pathologies such as lung cancer. However the long term risks particularly with regard to lung cancer are to date unclear. In the case of MOX the risk may indeed be different from that assigned to the individual components, plutonium and uranium. Several factors are influential (i) the dissolution of Pu depends on the physico-chemical properties, for example risk of lung cancer is increased 10 fold after Pu(NO 3 ) 2 as compared with PuO 2 . (ii) The solubility of Pu is variable whether delivered as PuO 2 or contained within MOX. (iii) The risk of cancer appears to increase with spatial homogeneity of the lung alpha dose. The objective of this study was to investigate the long term effects in rat lungs following MOX aerosol inhalation of similar particle size containing 2.5 or 7.1% Pu. Conscious rats were exposed to MOX aerosols using a 'nose-only' system and kept for their entire life (2-3 years). Different Initial Lung Deposits (ILDs) were obtained using different concentrations of the MOX suspension. Lung total alpha activity was determined in vivo at intervals over the study period by external counting as well as at autopsy in order to estimate the total lung dose. Anatomo-pathological and immunohistochemical analyses were performed on fixed lung tissue after euthanasia. The frequencies of lung pathologies and tumours were determined on lung sections at several different levels. In addition, autoradiography of lung sections was performed in order to assess the spatial localisation of a activity. Inhalation of MOX at ILD ranging from 1-20 kBq resulted in lung pathologies (90% of exposed rats

  20. Cesium migration in LMFBR fuel pins

    International Nuclear Information System (INIS)

    Karnesky, R.A.; Jost, J.W.; Stone, I.Z.

    1978-10-01

    The factors affecting the axial migration of cesium in mixed oxide fuel pins and the effects of cesium migration on fuel pin performance are examined. The development and application of a correlated model which will predict the occurrence of cesium migration in a mixed oxide (75 w/o UO 2 + 25 w/o PuO 2 ) fuel pins over a wide range of fabrication and irradiation conditions are described

  1. Correlation between fuel structure and mechanical properties of UO2

    International Nuclear Information System (INIS)

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

    1982-10-01

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

  2. Determination of hydrogen content in mixed oxide fuels (U,Pu)02 by three different methods: occluded gas, carrier gas at 9000C (apparatus Dupont de Nemours) and carrier gas at 17000C (apparatus Ithac 2000 Adamel). Operation, performance comparison of the three methods

    International Nuclear Information System (INIS)

    Lepetit; Lebrun; Haour; Dellasta; Vannier.

    1989-01-01

    Test conditions and performance are described. Results obtained allow a comparison and give the important parameters sampling conditions, sample storage and transfer, radiolysis and pyrohydrolysis. After a few hundred analysis fuels can be divided in two types: hydrogen content between 0.1 and 1 ppm corresponding to water fixation at the pellet surface, homogeneous content, not or slightly porous and with a perfect metallographic structure. Hydrogen content over 1 ppm, heterogeneous content from one sample to one another because of porosity and metallographic structure with cracks and cavities [fr

  3. Photocatalytic discoloration of reactive blue 5g dye in the presence of mixed oxides and with the addition of iron and silver

    International Nuclear Information System (INIS)

    Souza, M.C.P; Lenzi, G.G.; Jorge, L.M.M.; Santos, O.A.A.; Colpini, L.M.S.

    2011-01-01

    This work reports the use of cerium-titania-alumina-based systems modified with Ag and Fe by the wetness impregnation method for the discoloration of blue 5G dye. The techniques employed to characterize the photocatalysts were: temperature-programmed reduction (TPR), X-ray diffraction (XRD), specific surface area, average pore volume, and average pore diameter. The characterization results indicated that the photocatalysts had different crystalline structures and textural properties. Discoloration with the mixed oxide photocatalyst CeO 2 -TiO 2 -Al 2 O 3 gave a result similar to that of TiO 2 . On the other hand, the addition of Ag and Fe to the mixed oxide increased the discoloration and reaction rates of reactive blue 5G dyes. (author)

  4. Photocatalytic discoloration of reactive blue 5g dye in the presence of mixed oxides and with the addition of iron and silver

    Energy Technology Data Exchange (ETDEWEB)

    Souza, M.C.P; Lenzi, G.G.; Jorge, L.M.M.; Santos, O.A.A. [Universidade Estadual de Maringa (UEM), PR (Brazil). Dept. de Engenharia Quimica; Colpini, L.M.S. [Universidade Federal do Parana (UFPR), Palotina, PR (Brazil). Curso Superior de Tecnologia em Biocombustiveis

    2011-07-15

    This work reports the use of cerium-titania-alumina-based systems modified with Ag and Fe by the wetness impregnation method for the discoloration of blue 5G dye. The techniques employed to characterize the photocatalysts were: temperature-programmed reduction (TPR), X-ray diffraction (XRD), specific surface area, average pore volume, and average pore diameter. The characterization results indicated that the photocatalysts had different crystalline structures and textural properties. Discoloration with the mixed oxide photocatalyst CeO{sub 2}-TiO{sub 2}-Al{sub 2}O{sub 3} gave a result similar to that of TiO{sub 2}. On the other hand, the addition of Ag and Fe to the mixed oxide increased the discoloration and reaction rates of reactive blue 5G dyes. (author)

  5. Nanostructured Mn-Fe Binary Mixed Oxide: Synthesis, Characterization and Evaluation for Arsenic Removal.

    Science.gov (United States)

    Pillewan, Pradnya; Mukherjee, Shrabanti; Bansiwal, Amit; Rayalu, Sadhana

    2014-07-01

    Adsorption of arsenic on bimetallic Mn and Fe mixed oxide was carried out using both field as well as simulated water. The material was synthesized using hydrothermal method and characterized using X-ray diffraction (XRD) and scanning electron microscopy (SEM). Langmuir and Freundlich adsorption isotherms were computed using batch adsorption studies to determine the adsorption capacity of Mn-Fe binary mixed oxide for arsenic. Adsorption capacity for MFBMO obtained from Freundlich model was found to be 2.048 mg/g for simulated water and 1.084 mg/g for field water. Mn-Fe binary mixed oxide was found to be effective adsorbent for removal of arsenic from water.

  6. An overview of the effect of fuel properties on emissions from biomass fuels

    International Nuclear Information System (INIS)

    Graboski, M.S.

    1993-01-01

    Biofuels are considered to be environmentally benign since they are composed primarily of carbon, hydrogen and oxygen. The emissions resulting from biofuel use are dependent, however, on the system employed and how key fuel properties interact with the system. Two case studies are presented to demonstrate this fact. First, gasification and combustion of urban waste wood to produce electric power is investigated. Second, ethanol and ethanol derivatives are examined as reformulated gasoline additives

  7. Properties of U3Si2-Al dispersion fuel element and its application

    International Nuclear Information System (INIS)

    Yin Changgeng

    2001-01-01

    The properties of U 3 Si 2 fuel and U 3 Si 2 -Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U 3 Si 2 -Al dispersion fuel elements in the world are introduced

  8. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, T.

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

  9. Late-occurring pulmonary pathologies following inhalation of mixed oxide (uranium + plutonium oxide) aerosol in the rat.

    Science.gov (United States)

    Griffiths, N M; Van der Meeren, A; Fritsch, P; Abram, M-C; Bernaudin, J-F; Poncy, J L

    2010-09-01

    Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

  10. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  11. Tensile mechanical properties of U3Si2-Al fuel plate

    International Nuclear Information System (INIS)

    Xu Yong; Hu Huawei; Zhuang Hongquan; Wang Xishu

    2003-01-01

    The fuel plate made of fuel meat, with the U 3 Si 2 -Al dispersion fuel center, and 6061 Al alloy cladding, is a new kind of fuel used in research reactors. The mechanical property data of the fuel meat is the basic data in the design of fuel group, but the mechanical property of this fuel meat has not been studied all over the world till now. In this paper, the mechanical properties of U 3 Si 2 -Al fuel meats of different sizes used in research reactors are investigated and analyzed, and at the same time the carrying capacity of tensile in different directions are also compared. In order to get more knowledge about the mechanical properties of the fuel meat, the tensile experiment has been carried out repeatedly. Considering the lower ratio of elongation and the brittleness, the microscope has been used to examine the zone of fracture after tensile test. (authors)

  12. Perovskite solid electrolytes: Structure, transport properties and fuel cell applications

    DEFF Research Database (Denmark)

    Bonanos, N.; Knight, K.S.; Ellis, B.

    1995-01-01

    Doped barium cerate perovskites, first investigated by Iwahara and co-workers, have ionic conductivities of the order of 20 mS/cm at 800 degrees C making them attractive as fuel cell electrolytes for this temperature region. They have been used to construct laboratory scale fuel cells, which...... vapour transfer in a cell in which the perovskite is exposed to wet hydrogen on both sides. The evolution of transport properties with temperature is discussed in relation to structure. Neutron diffraction studies of doped and undoped barium cerate are reported, revealing a series of phase transitions...... between ambient temperature and 1000 degrees C. The available literature on chemical stability of cerate perovskites to reduction and attack by carbon dioxide is reviewed in brief....

  13. Mechanical properties of fuel debris for defueling toward decommissioning

    International Nuclear Information System (INIS)

    Hoshino, Takanori; Kitagaki, Toru; Yano, Kimihiko; Okamura, Nobuo; Koizumi, Kenji; Ohara, Hiroshi; Fukasawa, Tetsuo

    2015-01-01

    In the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F), safe and steady defueling work is required. Before defueling 1F, it is necessary to evaluate fuel debris for properties related to the defueling procedure and technology. While defueling after the Three Mile Island Nuclear Power Plant Unit 2 (TMI-2) accident, a core boring system played an important role. Considering the working principle of core boring, hardness, elastic modulus, and fracture toughness were found to be important fuel debris properties that had a profound effect on the performance of the boring machine. It is speculated that uranium and zirconium oxide solid solution ((U,Zr)O_2) is one of the major materials of fuel debris in 1F, according to the TMI-2 accident experience and the results of past severe accident studies. In addition, the Zr content of 1F fuel debris is expected to be higher than that of TMI-2 debris, because the 1F reactors were boiling-water reactor (BWR). In this report, the mechanical properties of (U,Zr)O_2 are evaluated in the ZrO_2 content range from 10% to 65%. The hardness, elastic modulus, and fracture toughness were measured by Vickers test, ultrasonic pulse echo method, and indentation fracture method, respectively. In the ZrO_2 content range under 50%, the Vickers hardness and fracture toughness of (U,Zr)O_2 increased, and the elastic modulus decreased slightly with ZrO_2 content. In the case of 55% and 65% ZrO_2, all of those measures increased slightly with ZrO_2 content. Summarizing those results, ZrO_2 content affects mechanical properties significantly in the case of low ZrO_2 content. Higher Zr content (exceeding 50%) has little effect on mechanical properties. In the future, nonradioactive surrogate debris will be necessary for small-scale functional and large-scale mockup tests of various defueling technologies. These results are useful to select the material for surrogate debris. (author)

  14. Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides.

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C; Robisson, Anne-Charlotte; Lebreton, Florent; Aufore, Laurence; Scheinost, Andreas C; Martin, Philippe M

    2016-03-07

    Innovative americium-bearing uranium-plutonium mixed oxides U1-yPuyO2-x are envisioned as nuclear fuel for sodium-cooled fast neutron reactors (SFRs). The oxygen-to-metal (O/M) ratio, directly related to the oxidation state of cations, affects many of the fuel properties. Thus, a thorough knowledge of its variation with the sintering conditions is essential. The aim of this work is to follow the oxidation state of uranium, plutonium, and americium, and so the O/M ratio, in U0.750Pu0.246Am0.004O2-x samples sintered for 4 h at 2023 K in various Ar + 5% H2 + z vpm H2O (z = ∼ 15, ∼ 90, and ∼ 200) gas mixtures. The O/M ratios were determined by gravimetry, XAS, and XRD and evidenced a partial oxidation of the samples at room temperature. Finally, by comparing XANES and EXAFS results to that of a previous study, we demonstrate that the presence of uranium does not influence the interactions between americium and plutonium and that the differences in the O/M ratio between the investigated conditions is controlled by the reduction of plutonium. We also discuss the role of the homogeneity of cation distribution, as determined by EPMA, on the mechanisms involved in the reduction process.

  15. On the effects of fuel properties and injection timing in partially premixed compression ignition of low octane fuels

    KAUST Repository

    Naser, Nimal

    2017-06-29

    A better understanding on the effects of fuel properties and injection timing is required to improve the performance of advanced engines based on low temperature combustion concepts. In this work, an experimental and computational study was conducted to investigate the effects of physical and chemical kinetic properties of low octane fuels and their surrogates in partially premixed compression ignition (PPCI) engines. The main objective was to identify the relative importance of physical versus chemical kinetic properties in predicting practical fuel combustion behavior across a range of injection timings. Two fuel/surrogate pairs were chosen for comparison: light naphtha (LN) versus the primary reference fuel (PRF) with research octane number of 65 (PRF 65), and FACE (fuels for advanced combustion engines) I gasoline versus PRF 70. Two sets of parametric studies were conducted: the first varied the amount of injected fuel mass at different injection timings to match a fixed combustion phasing, and the second maintained the same injected fuel mass at each injection timing to assess resulting combustion phasing changes. Full-cycle computational fluid dynamic engine simulations were conducted by accounting for differences in the physical properties of the original and surrogate fuels, while employing identical chemical kinetics. The simulations were found to capture trends observed in the experiments, while providing details on spatial mixing and chemical reactivity for different fuels and injection timings. It was found that differences in physical properties become increasingly important as injection timing was progressively delayed from premixed conditions, and this was rationalized by analysis of mixture stratification patterns resulting from injection of fuels with different physical properties. The results suggest that accurate descriptions of both physical and chemical behavior of fuels are critical in predictive simulations of PPCI engines for a wide range of

  16. On the effects of fuel properties and injection timing in partially premixed compression ignition of low octane fuels

    KAUST Repository

    Naser, Nimal; Jaasim, Mohammed; Atef, Nour; Chung, Suk-Ho; Im, Hong G.; Sarathy, Mani

    2017-01-01

    A better understanding on the effects of fuel properties and injection timing is required to improve the performance of advanced engines based on low temperature combustion concepts. In this work, an experimental and computational study was conducted to investigate the effects of physical and chemical kinetic properties of low octane fuels and their surrogates in partially premixed compression ignition (PPCI) engines. The main objective was to identify the relative importance of physical versus chemical kinetic properties in predicting practical fuel combustion behavior across a range of injection timings. Two fuel/surrogate pairs were chosen for comparison: light naphtha (LN) versus the primary reference fuel (PRF) with research octane number of 65 (PRF 65), and FACE (fuels for advanced combustion engines) I gasoline versus PRF 70. Two sets of parametric studies were conducted: the first varied the amount of injected fuel mass at different injection timings to match a fixed combustion phasing, and the second maintained the same injected fuel mass at each injection timing to assess resulting combustion phasing changes. Full-cycle computational fluid dynamic engine simulations were conducted by accounting for differences in the physical properties of the original and surrogate fuels, while employing identical chemical kinetics. The simulations were found to capture trends observed in the experiments, while providing details on spatial mixing and chemical reactivity for different fuels and injection timings. It was found that differences in physical properties become increasingly important as injection timing was progressively delayed from premixed conditions, and this was rationalized by analysis of mixture stratification patterns resulting from injection of fuels with different physical properties. The results suggest that accurate descriptions of both physical and chemical behavior of fuels are critical in predictive simulations of PPCI engines for a wide range of

  17. Thermal diffusivity and conductivity of thorium- uranium mixed oxides

    Science.gov (United States)

    Saoudi, M.; Staicu, D.; Mouris, J.; Bergeron, A.; Hamilton, H.; Naji, M.; Freis, D.; Cologna, M.

    2018-03-01

    Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O2 (0 ≤ x ≤ 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th,U)O2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th,U)O2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems.

  18. Mixed oxides forming in CdO-OsO2 and PbO-OsO2 systems

    International Nuclear Information System (INIS)

    Shaplygin, I.S.; Lazarev, V.B.

    1978-01-01

    The formation of mixed oxides in systems CdO-OsO 2 and PbO-OsO 2 was studied by derivatography, X-ray phase analysis, and IR spectroscopy. It was established that in the system CdO-OsO 2 cubic phases of Cdsub(2)Ossub(2)Osub(6.7) and Cdsub(2)Ossub(2)Osub(7.2) with a pyrochlor-type structure, as well as new compounds with the structure of rhombic perovskite CdOsO 3 and Cd(Cdsub(0.5)Ossub(0.5))Osub(3) containing osmium in the formal degree of oxidation +4 and +6, respectively, are formed. The thermal stability of the compounds obtained was studied and their electrical properties determined

  19. Comparison of the redox activities of sol-gel and conventionally prepared Bi-Mo-Ti mixed oxides

    Energy Technology Data Exchange (ETDEWEB)

    Wildberger, M.; Grundwaldt, J.D.; Mallat, T.; Baiker, A. [Lab. of Technical Chemistry, Swiss Federal Inst. of Technology, ETH-Zentrum, Zuerich (Switzerland)

    1998-12-31

    Novel sol-gel Bi-Mo-Ti oxides have been prepared and characterized by XRD, XPS, FT-Raman and HRTEM. The surface Bi{sup 3+} and Mo{sup 6+} species of some xerogels and an aerogel could be reduced and oxidized at room temperature, whereas the conventionally prepared reference materials were not reduced by H{sub 2} below 300 C. The unusual redox properties, under very mild conditions, are likely due to the unique morphology of Bi-Mo-oxides stabilized by titania. During butadiene oxidation to furan at above 400 C to sol-gel mixed oxides restructured considerably and their performance was barely better than that of titania-supported Bi-Mo oxides. (orig.)

  20. Ni–Ta–O mixed oxide catalysts for the low temperature oxidative dehydrogenation of ethane to ethylene

    KAUST Repository

    Zhu, Haibo

    2015-09-01

    The "wet" sol-gel and "dry" solid-state methods were used to prepare Ni-Ta-O mixed oxide catalysts. The resulting Ni-Ta oxides exhibit high activity and selectivity for the low temperature oxidative dehydrogenation of ethane to ethylene. The Ta/(Ni + Ta) atomic ratios (varying from 0 to 0.11 in "wet" sol-gel method, and from 0 to 0.20 in "dry" solid-state method) as well as the preparation methods used in the synthesis, play important roles in controlling catalyst structure, activity, selectivity and stability in the oxidative dehydrogenation of ethane. Electron microscopy characterizations (TEM, EELS mapping, and HAADF-STEM) clearly demonstrate that the Ta atoms are inserted into NiO crystal lattice, resulting in the formation of a new Ni-Ta oxide solid solution. More Ta atoms are found to be located at the lattice sites of crystal surface in sol-gel catalyst. While, a small amount of thin layer of Ta2O5 clusters are detected in solid-state catalyst. Further characterization by XRD, N2 adsorption, SEM, H2-TPR, XPS, and Raman techniques reveal different properties of these two Ni-Ta oxides. Due to the different properties of the Ni-Ta oxide catalysts prepared by two distinct approaches, they exhibit different catalytic behaviors in the ethane oxidative dehydrogenation reaction at low temperature. Thus, the catalytic performance of Ni-Ta-O mixed oxide catalysts can be systematically modified and tuned by selecting a suitable synthesis method, and then varying the Ta content. ©2015 Elsevier Inc. All rights reserved.

  1. Co-Mn-Al Mixed Oxides as Catalysts for Ammonia Oxidation to N2O.

    Czech Academy of Sciences Publication Activity Database

    Ludvíková, Jana; Jablońska, M.; Jirátová, Květa; Chmielarz, L.; Balabánová, Jana; Kovanda, F.; Obalová, L.

    2016-01-01

    Roč. 42, č. 3 (2016), s. 2669-2690 ISSN 0922-6168 R&D Projects: GA ČR GA14-13750S Institutional support: RVO:67985858 Keywords : Co-Mn-Al mixed oxide s * catalytic ammonia oxidation * N2O production * mechanochemical production Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 1.369, year: 2016

  2. Degradation of organophosphorus pesticide parathion methyl on nanostructured titania-iron mixed oxides

    Energy Technology Data Exchange (ETDEWEB)

    Henych, Jiří, E-mail: henych@iic.cas.cz [Department of Material Chemistry, Institute of Inorganic Chemistry AS CR v.v.i., 25068 Řež (Czech Republic); Štengl, Václav; Slušná, Michaela; Matys Grygar, Tomáš [Department of Material Chemistry, Institute of Inorganic Chemistry AS CR v.v.i., 25068 Řež (Czech Republic); Janoš, Pavel; Kuráň, Pavel; Štastný, Martin [Faculty of the Environment, J.E. Purkyně University, Králova Výšina 7, 400 96 Ústí nad Labem (Czech Republic)

    2015-07-30

    Highlights: • Ti–Fe mixed oxides were synthesized via low-temperature one-pot method. • Mixed oxides were used for degradation of parathion methyl. • Pure reference oxide samples showed no degradation ability. • Mixed oxides reached 70% degree of conversion of parathion methyl. - Abstract: Titania-iron mixed oxides with various Ti:Fe ratio were prepared by homogeneous hydrolysis of aqueous solutions of titanium(IV) oxysulphate and iron(III) sulphate with urea as a precipitating agent. The synthesized samples were characterized by X-ray diffraction, Raman and infrared spectroscopy, scanning and transmission electron microscopy, XRF analysis, specific surface area (BET) and porosity determination (BJH). These oxides were used for degradation of organophosporus pesticide parathion methyl. The highest degradation efficiency approaching <70% was found for the samples with Ti:Fe ratio 0.25:1 and 1:0.25. Contrary, parathion methyl was not degraded on the surfaces of pure oxides. In general, the highest degradation rate exhibited samples consisted of the iron or titanium oxide containing a moderate amount of the admixture. However, distinct correlations between the degradation rate and the sorbent composition were not identified.

  3. SIMMER-III analytic thermophysical property model

    International Nuclear Information System (INIS)

    Morita, K; Tobita, Y.; Kondo, Sa.; Fischer, E.A.

    1999-05-01

    An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms are designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless steel, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-core materials. (author)

  4. Impact of physical properties of mixture of diesel and biodiesel fuels on hydrodynamic characteristics of fuel injection system

    Directory of Open Access Journals (Sweden)

    Filipović Ivan M.

    2014-01-01

    Full Text Available One of the alternative fuels, originating from renewable sources, is biodiesel fuel, which is introduced in diesel engines without major construction modifications on the engine. Biodiesel fuel, by its physical and chemical properties, is different from diesel fuel. Therefore, it is expected that by the application of a biodiesel fuel, the characteristic parameters of the injection system will change. These parameters have a direct impact on the process of fuel dispersion into the engine cylinder, and mixing with the air, which results in an impact on the quality of the combustion process. Method of preparation of the air-fuel mixture and the quality of the combustion process directly affect the efficiency of the engine and the level of pollutant emissions in the exhaust gas, which today is the most important criterion for assessing the quality of the engine. The paper presents a detailed analysis of the influence of physical properties of a mixture of diesel and biodiesel fuels on the output characteristics of the fuel injection system. The following parameters are shown: injection pressure, injection rate, the beginning and duration of injection, transformation of potential into kinetic energy of fuel and increase of energy losses in fuel injection system of various mixtures of diesel and biodiesel fuels. For the analysis of the results a self-developed computer program was used to simulate the injection process in the system. Computational results are verified using the experiment, for a few mixtures of diesel and biodiesel fuels. This paper presents the verification results for diesel fuel and biodiesel fuel in particular.

  5. Basic properties of a zirconia based fuel material for LWRs

    International Nuclear Information System (INIS)

    Degueldre, C.; Paratte, J.M.

    1997-01-01

    The properties of zirconia cubic solid solutions doped with yttria, erbia and ceria or thoria are investigated with emphasis on the potential use of this material as inert matrix fuel for plutonium incineration in a light water reactor (LWR). The material is selected on the basis of its neutronic properties. Zr and Y are not neutron absorbers. Among the rare earth elements, Er was identified as a suitable burnable poison. The high density cubic solid solution is stable for a rather large range of compositions and from room temperature up to about 3000 K. Samples irradiated under low and high energy Xe ion irradiation up to a fluence of 1.8.10 16 Xe.cm -2 were investigated by transmission electron microscopy. Low energy (60 keV) Xe ions did not produce amorphization. From the observed bubble formation, swelling values during irradiation at room temperature or at high temperature (925 K) were estimated to be 0.1-0.72% by volume. Furthermore, no amorphization was obtained by Xe irradiation under extreme conditions such as high energy (1.5 MeV) Xe ion irradiation and low temperature (20 K). This confirms the robustness of this material and argues in favour of the selection of a zirconia based material as an advanced nuclear fuel for plutonium incineration. (author) 5 figs., 1 tab., 17 refs

  6. TSUNAMI analysis of the applicability of proposed experiments to reactor-grade and weapons-grade mixed-oxide systems

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Hopper, Calvin M.; Elam, Karla R.

    2005-01-01

    The applicability of proposed critical experiments for the criticality code validation of a series of prototypic reactor-grade and weapons-grade mixed-oxide systems has been assessed with the TSUNAMI methodology from SCALE 5. The application systems were proposed by the Nuclear Energy Agency (NEA) Organization for Economic Cooperation and Development (OECD) Working Party on Nuclear Criticality Safety MOX Experimental Needs Working Group. Forty-eight application systems were conceived to envelope the range of conditions in processing and fabrication of reactor-grade and weapons-grade MOX fuel. The applicability of 303 existing critical benchmarks to each of the 48 applications was assessed, and validation coverage was found to be lacking for certain applications. Two series of proposed critical experiments were also considered in this analysis. The TSUNAMI analysis has revealed that both series of proposed experiments are applicable to numerous configurations of the reactor-grade and weapons-grade systems. A detailed assessment of which experiments were revealed by TSUNAMI to be most applicable to specific prototypic fuel processing systems has been performed. (author)

  7. Investigation of a piezoelectric droplet delivery method for fuel injection and physical property evaluation

    Science.gov (United States)

    Zhao, Wei; Menon, Shyam

    2017-11-01

    A piezoelectric droplet generator is investigated to deliver liquid hydrocarbon fuels to a micro-combustor application. Besides fuel delivery, the setup is intended to measure fuel physical properties such as viscosity and surface tension. These properties are highly relevant to spray generation in internal combustion engines. Accordingly, a drop-on-demand piezoelectric dispenser is used to generate fuel droplet trains, which are studied using imaging and Phase Doppler Particle Anemometry (PDPA). The diagnostics provide information regarding droplet size and velocity and their evolution over time. The measurements are correlated with results from one-dimensional (1D) models that incorporate sub-models for piezo-electric actuation and droplet vaporization. By validating the 1D models for fuels with known physical properties, a technique is developed that has the capability to meter low-vapor pressure liquid fuels to the microcombustor and use information from the droplet train to calculate physical properties of novel fuels.

  8. Models for MOX fuel behaviour. A selective review

    International Nuclear Information System (INIS)

    Massih, Ali R.

    2006-01-01

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO 2 fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO 2 . In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO 2 fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO 2 fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO 2 vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO 2 . This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation

  9. Models for MOX fuel behaviour. A selective review

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2006-12-15

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO{sub 2} fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO{sub 2}. In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO{sub 2} fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO{sub 2} fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO{sub 2} vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO{sub 2}. This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation.

  10. Evaluation of spent fuel properties from a conceptual PEACER core

    International Nuclear Information System (INIS)

    Lim, Jae Yong; Kim, Myung Hyun; Kim, Chang Hyo; Hwang, Il Soon

    2003-01-01

    In this paper, a new conceptual core design, PEACER was evaluated in aspect of core performance and spent fuel properties. The core shape is like a pancake to increase axial neutron leakage. Square lattice array was applied which was suitable to decrease the flow speed of Pb-Bi coolant. Although over 30% TRU produced by pyroprocessing was loaded in U-Zr metal fuel, the cycle length of 1 year was achieved and the relative assembly power peaking was less than 1.3. In order to confirm nuclear performance of PEACER core design, several performance indices were adopted and developed. Simple indices such as FIR and FG were used to evaluate fissile breeding. BCM, TG, SNS, and OR calculated by plutonium composition vectors were chosen to distinguish the competency of proliferation resistance. For the estimation of transmutation capability, D-value and extended effective fission half-life time(T EX ) were used. According to these indices, the PEACER core had the better performance compared with other conventional reactor cores although fissile breeding was not acquired

  11. US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

    2016-03-01

    The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

  12. Thermophysical properties of the irradiated uranium-zirconium fuel

    International Nuclear Information System (INIS)

    Gajduchenko, A.B.

    2008-01-01

    The dependence of the thermophysical properties of metallic nuclear fuel, i.e. Zr alloy 40U, in a wide temperature range as a function of accumulated fission products amount is presented. Both non-irradiated and irradiated test pieces with different degrees of accumulation of fission products, i.e. 0.4, 0.6, and 0.9 g/cm 3 , are investigated. The specific heat is measured in the range of 50-1000 deg C, the thermal diffusivity is measured in the range 300-1000 deg C, and the variation of the dimensions and density of the samples on heating is also investigated. The thermal conductivity in the range of 50-1000 deg C is calculated on the basis of the experimental data [ru

  13. Innovative nuclear fuels and applications. Part 1: limits of today's fuels and concepts for innovative fuels. Part 2: materials properties, irradiation performance and gaps in our knowledge

    International Nuclear Information System (INIS)

    Matzke, H.

    2000-01-01

    Part I of this contribution on innovative nuclear fuels gives a summary of current developments and problems of today's fuels, i.e. enriched UO 2 and UO 2 with a few % of PUO 2 (MOX fuel) or Gd 2 O 3 (as burnable neutron poison). The problems and property changes caused by high burnups (e.g. degradation of the thermal conductivity, polygonization or formation of the rim-structure) are discussed. Subsequently, the concepts for new fuels to burn excess Pu and to achieve an effective transmutation of the minor actinides Np, Am and Cm are treated. The criteria for the choice of suitable fuels and different fuel types (high Pu-content fuels, nitrides, U-free fuels, inert matrix supported fuels, cercers, cermets, etc.) are discussed. Part II of this contribution on innovative nuclear fuels deals with the properties of relevance of the different materials suggested to be used in innovative fuels which range from pure actinide fuel such as PuN and AmO 2 to spinel MgAl 2 O 4 and zircon ZrSiO 4 for inert matrix-based fuels, etc. The available knowledge on materials research aspects is summarized with emphasis on the physics of radiation damage. It is shown that significant gaps in the present knowledge exist, e.g. for the minor actinide compounds, and suggestions are made to fill these gaps in order to achieve a sufficient data base to design and operate suitable innovative fuels in a near future. (author)

  14. Synthesis of mesoporous cerium-zirconium mixed oxides by hydrothermal templating method

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Mesoporous cerium-zirconium mixed oxides were prepared by hydrothermal method using cetyl trimethyl ammonium bromide (CTAB) as template.The effects of amount of template,pH value of solution and hydrothermal temperature on mesostructure of samples were systematically investigated.The final products were characterized by XRD,TEM,FT-IR,and BET.The results indicate that all the cerium-zirconium mixed oxides present a meso-structure.At molar ratio of n(CTAB)/n((Ce)+(Zr))=0.15,pH value of 9,and hydrothermal temperature of 120 ℃,the samples obtained possess a specific surface area of 207.9 m2/g with pore diameter of 3.70 nm and pore volume of 0.19 cm3/g.

  15. Formation of peripheral porosity regions around urania in zirconia-urania mixed oxide powder compact sintering

    International Nuclear Information System (INIS)

    Das, P.; Choudhury, R.

    1992-01-01

    Sintering studies of zirconia-urania mixed oxide powder compacts (in stages of 5% urania up to a maximum of 20% addition) were carried out at temperatures between 1000-1400deg C for various soaking periods. The formation of a peripheral porosity region around comparatively coarser urania particle was a characteristic feature in this mixed oxide sintered compact. At even a higher sintering temperature (1800deg C), where extensive solid solution formation takes place, this porosity region demarcates the solutionized particles from the host zirconia apparently acting as a discontinuity in the system. Relative shrinkage difference between the dissimilar particles probably contributes to the porosity regions around the minor second phase at a lower temperature while at higher temperature generation of 'Kirkendall porosity' may be responsible for such an effect. (orig.)

  16. Neutronics benchmarks of mixed-oxide fuels using the SCALE/CENTRM sequence

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Fox, P.B.

    2000-01-01

    The purpose of this study is to determine and document the reactor physics parameters (multiplication factors, spatially dependent flux ratios, and spacially dependent reaction rates ) for several distinct sets of problems using two distinct resonance cross-section processing techniques. In SCALE, by default, resonances are processed using NITAWL, which utilizes the Nordheim Integral Treatment. The results produced using this sequence are considered to be the base results. A second set of results are produced by replacing NITAWL with CENTRM/PMC. CENTRM produces point-wise fluxes for a given geometry configuration and set of isotopes. Using these fluxes, PMC produces problem-dependent self-shielding cross sections. Both sequences use ENDF/B-V cross-section data

  17. Safeguards implementation practices for a model mixed oxide recycle fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Glancy, J.E.

    1976-05-01

    Conclusions on the magnitude of the inspection effort and its effectiveness in meeting IAEA objectives are summarized: the minimum goal quantity that can be detected with 95 percent probability is 6 kg Pu. eight to nine resident inspectors are necessary to monitor all operations (3 shifts a day) and to perform verification measurements (23 per day). The facility must also perform inventories monthly if the goal of 6.0 kg is to be attained. Bi-monthly inventory and continual on-site resident inspection can achieve a detection goal quantity of 8.0 kg Pu with 95 percent detection probability. For annual inventories, the detection goal quantity for a 95 percent detection probability is 40 kg Pu. The probability for detecting 8 kg is less than 20 percent. The total inspection effort for a 200 MT/year RFP is estimated to exceed 15 man-years if IAEA headquarters support and analytical laboratory time are added to on-site manpower requirements. Surveillance and containment in the form of seals must be applied to prevent double inventorying and to assure that samples are extracted from the vessel or container being measured. These assurances will probably be achieved by on-line instrumentation installed by the IAEA prior to plant start-up. For assessment, detailed knowledge of plant equipment and floor layout including piping used to transport material is needed. Such information should be requested from the facility prior to preparation of the SIP. Present NDA appears capable of measuring all material in the RFP; however, verification of residual hold-up will require development of new procedures. In-process inventory verification will require special procedures but can be accomplished with present technology. Alternatives to accountability such as perimeter surveillance and containment of declared material within the perimeter should be studied in order to improve detection sensitivity or decrease the effort. (DLC)

  18. External radiation exposure and radiotoxicity considerations in plutonium/uranium mixed-oxide fuel fabrication

    International Nuclear Information System (INIS)

    Williams, R.A.; Crosby, E.H.

    1974-01-01

    Nuclear-reactor-produced plutonium emits significant electromagnetic and neutron radiations. In addition, because of its high specific alpha activity and its tendency to deposit in the lung and the soft tissues of the bone, plutonium presents a significant radiotoxicity hazard. Shielding, containment, and dosimetry techniques practiced at the Nuclear Materials and Equipment Corporation (NUMEC) have resulted in exemplary annual safety report statistics relevant to non-remote-handling plutonium operations. Whereas a few employees exceeded the NUMEC external exposure action level, no employee exceeded the Regulatory maximum permissible external exposure. In addition, a few employees were observed to have a lung burden in excess of the minimum sensitivity of an in vivo counting system, and one employee was observed by in vivo counting to have a lung burden in excess of the maximum permissible for a brief period. No employee was observed to have a body burden as indicated by a positive quarterly urinalysis result. Further, there were no serious incidents at the facility requiring immediate Regulatory notification, and there were no moderate incidents at the facility requiring 24-h Regulatory notification. However, there were a few reportable incidents at the facility requiring 30-day Regulatory notification, and there were a few minor incidents at the facility requiring the preparation of a NUMEC Incident Report. Details of this safety record are presented along with the health physics techniques that have contributed to the results

  19. Safeguards implementation practices for a model mixed oxide recycle fuel fabrication facility

    International Nuclear Information System (INIS)

    Glancy, J.E.

    1976-05-01

    Conclusions on the magnitude of the inspection effort and its effectiveness in meeting IAEA objectives are summarized: the minimum goal quantity that can be detected with 95 percent probability is 6 kg Pu. eight to nine resident inspectors are necessary to monitor all operations (3 shifts a day) and to perform verification measurements (23 per day). The facility must also perform inventories monthly if the goal of 6.0 kg is to be attained. Bi-monthly inventory and continual on-site resident inspection can achieve a detection goal quantity of 8.0 kg Pu with 95 percent detection probability. For annual inventories, the detection goal quantity for a 95 percent detection probability is 40 kg Pu. The probability for detecting 8 kg is less than 20 percent. The total inspection effort for a 200 MT/year RFP is estimated to exceed 15 man-years if IAEA headquarters support and analytical laboratory time are added to on-site manpower requirements. Surveillance and containment in the form of seals must be applied to prevent double inventorying and to assure that samples are extracted from the vessel or container being measured. These assurances will probably be achieved by on-line instrumentation installed by the IAEA prior to plant start-up. For assessment, detailed knowledge of plant equipment and floor layout including piping used to transport material is needed. Such information should be requested from the facility prior to preparation of the SIP. Present NDA appears capable of measuring all material in the RFP; however, verification of residual hold-up will require development of new procedures. In-process inventory verification will require special procedures but can be accomplished with present technology. Alternatives to accountability such as perimeter surveillance and containment of declared material within the perimeter should be studied in order to improve detection sensitivity or decrease the effort

  20. Summary report of phase I residual holdup measurements for a mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Woodsum, H.C.

    1978-03-01

    Metal surface-powder adherence tests showed that the average mean values for direct impingement was 60 to 80 g/ft 2 , whereas the average mean values of the colloidal samples were 0.2 to 2.5 g/ft 2 . Thus, it is advantageous to design powder processing equipment to reduce direct impingement wherever possible. Holdup of powder appears to be relatively independent of the surface material or finish, and it is reduced significantly by low-frequency vibration of the surface. Under colloidal conditions, ThO 2 produces more residual material than UO 2 and is preferentially deposited from a UO 2 --ThO 2 blend. Pure ThO 2 and high enrichment blends of ThO 2 in UO 2 are expected to produce a significant, persistent electrostatic charge, thus increasing residual holdup. Residual holdup in the clean scrap recovery system (CSRS) could be reduced by 25%. Comparison of CSRS holdup and powder adherence-metal surface data indicated that the areal density of residual material (40 g/ft 2 ) was considerably higher than for colloidal suspension ( 2 ). Steady-state residual holdup factor for sintered-metal filters was 13 g/ft 2 of filter surface area under optimum conditions. During the pellet grinding tests, residual material built up in the system at rate of about 100 g/h to an estimated limit of 1.4 kg, primarily within the particle collector shroud. During dry grinding, 97% of this residue was contained within the shroud, and during wet grinding only 50% was contained in the shroud owing to inertial effects of the rotating wheel and water coolant

  1. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  2. Standard recommended practice for examination of fuel element cladding including the determination of the mechanical properties

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Guidelines are provided for the post-irradiation examination of fuel cladding and to achieve better correlation and interpretation of the data in the field of radiation effects. The recommended practice is applicable to metal cladding of all types of fuel elements. The tests cited are suitable for determining mechanical properties of the fuel elements cladding. Various ASTM standards and test methods are cited

  3. Study of the solid solution formation in mixed oxides by X-ray diffraction

    International Nuclear Information System (INIS)

    Riella, H.G.

    1984-01-01

    A method to determine the plutonium distribution in mixed oxides - UO 2 /PuO 2 is described. The distribution function and the medium size of crystallite are obtained from the X-ray diffraction profile. Through the deconvolution by Fourier analysis, the X-ray diffraction profile is obtained without the influence of the difractrometer. Some experimental results for different samples of UO 2 -PuO 2 discussed. (Author) [pt

  4. Alkali Metals as Promoters in Co-Mn-Al Mixed Oxide for N2O Decomposition

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Karásková, K.; Wach, A.; Kustrowski, P.; Mamulová-Kutláková, K.; Michalik, S.; Jirátová, Květa

    462-463, JUL 10 (2013), s. 227-235 ISSN 0166-9834 R&D Projects: GA TA ČR TA01020336 Grant - others:MŠMT(CZ) CZ.1.05/2.1.00/03/0100; MŠMT(CZ) CZ.1.07/2.3.00/20.0074 Institutional support: RVO:67985858 Keywords : layered double hydroxides * hydrothermal reaction * mixed oxides * supported catalysts * ethanol total oxidation Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  5. K‑Doped Co−Mn−Al Mixed Oxide Catalyst for N2O Abatement from\

    Czech Academy of Sciences Publication Activity Database

    Pacultová, K.; Karásková, K.; Kovanda, F.; Jirátová, Květa; Šrámek, J.; Kustrovski, P.; Kotarba, A.; Chromčáková, Ž.; Kočí, K.; Obalová, L.

    2016-01-01

    Roč. 55, č. 26 (2016), s. 7076-7084 ISSN 0888-5885 R&D Projects: GA ČR GA14-13750S; GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : Co-Mn-Al mixed oxide * N2O decomposition * HNO3 pilot plant Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.843, year: 2016

  6. Impact of uranium-233/thorium cycle on advanced accountability concepts and fabrication facilities. Addendum 2 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Crandall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to cover the possible fabrication of uranium-233/thorium fuels. Revisions to Phase II of the DYMAC plan which would be necessitated by such a process are specified. These revisions include shielding requirements, measurement systems, licensing conditions, and safeguards considerations. The impact of the uranium/thorium cycle on a large-scale fuel fabrication facility was also reviewed; it was concluded that the essentially higher radioactivity of uranium/thorium feeds would lead to increased difficulties which tend to preclude early commercial application of the process. An amended schedule for Phase II is included

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  8. Diesel Surrogate Fuels for Engine Testing and Chemical-Kinetic Modeling: Compositions and Properties.

    Science.gov (United States)

    Mueller, Charles J; Cannella, William J; Bays, J Timothy; Bruno, Thomas J; DeFabio, Kathy; Dettman, Heather D; Gieleciak, Rafal M; Huber, Marcia L; Kweon, Chol-Bum; McConnell, Steven S; Pitz, William J; Ratcliff, Matthew A

    2016-02-18

    The primary objectives of this work were to formulate, blend, and characterize a set of four ultralow-sulfur diesel surrogate fuels in quantities sufficient to enable their study in single-cylinder-engine and combustion-vessel experiments. The surrogate fuels feature increasing levels of compositional accuracy (i.e., increasing exactness in matching hydrocarbon structural characteristics) relative to the single target diesel fuel upon which the surrogate fuels are based. This approach was taken to assist in determining the minimum level of surrogate-fuel compositional accuracy that is required to adequately emulate the performance characteristics of the target fuel under different combustion modes. For each of the four surrogate fuels, an approximately 30 L batch was blended, and a number of the physical and chemical properties were measured. This work documents the surrogate-fuel creation process and the results of the property measurements.

  9. Influence of Fatty Acid Methyl Esters on Fuel properties of Biodiesel ...

    African Journals Online (AJOL)

    Influence of Fatty Acid Methyl Esters on Fuel properties of Biodiesel Produced from the. Seeds Oil of Curcubita ... Gas chromatograph coupled with mass spectrophotometer (GC-MS). The results indicate ..... Chemical and physical properties of ...

  10. Macrophages as key elements of Mixed-oxide [U-Pu(O2)] distribution and pulmonary damage after inhalation?

    Science.gov (United States)

    Van der Meeren, Anne; Moureau, Agnes; Griffiths, Nina M

    2014-11-01

    Abstract Purpose: To investigate the consequences of alveolar macrophage (AM) depletion on Mixed OXide fuel (MOX: U, Pu oxide) distribution and clearance, as well as lung damage following MOX inhalation. Rats were exposed to MOX by nose only inhalation. AM were depleted with intratracheal administration of liposomal clodronate at 6 weeks. Lung changes, macrophage activation, as well as local and systemic actinide distribution were studied up to 3 months post-inhalation. Clodronate administration modified excretion/retention patterns of α activity. At 3 months post-inhalation lung retention was higher in clodronate-treated rats compared to Phosphate Buffered Saline (PBS)-treated rats, and AM-associated α activity was also increased. Retention in liver was higher in clodronate-treated rats and fecal and urinary excretions were lower. Three months after inhalation, rats exhibited lung fibrotic lesions and alveolitis, with no marked differences between the two groups. Foamy macrophages of M2 subtype [inducible Nitric Oxide Synthase (iNOS) negative but galectin-3 positive] were frequently observed, in correlation with the accumulation of MOX particles. AM from all MOX-exposed rats showed increased chemokine levels as compared to sham controls. Despite the transient reduced AM numbers in clodronate-treated animals no major differences on lung damage were observed as compared to non-treated rats after MOX inhalation. The higher lung activity retention in rats receiving clodronate seems to be part of a general inflammatory response and needs further investigation.

  11. The transportation of PuO2 and MOX fuel and management of irradiated MOX fuel

    International Nuclear Information System (INIS)

    Dyck, H.P.; Rawl, R.; Durpel, L. van den

    2000-01-01

    Information is given on the transportation of PuO 2 and mixed-oxide (MOX) fuel, the regulatory requirements for transportation, the packages used and the security provisions for transports. The experience with and management of irradiated MOX fuel and the reprocessing of MOX fuel are described. Information on the amount of MOX fuel irradiated is provided. (author)

  12. A study on coated particle fuel properties and performances and phase-I data base establishment

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Hyo Cheol; Im, Byeong Ju; Yun, Sang Pil; Son, Seung Beom; Lee, Gyeong Hui; Jang, Jeong Nam

    2006-03-01

    For the successful development of the high temperature gas cooled reactor acquisition and generation of the high temperature properties of reactor materials, especially temperature and burn-up dependent properties of coated particle fuel and fuel element, are crucially essential. Recently national project for HTGR for hydrogen production has been kicked off. However, we have had little experience on this new challenges. Therefore, it became necessary to build up the materials properties and fuel performance data base. In this study, a primitive properties and performance DB for coated particle fuel was developed. This database report consists two sections: materials properties and fuel performance. The materials properties has three parts: kernel materials, carbide coating materials, and fuel elements and graphite matrix. UO 2 and UCO belong to kernel materials while PyC, SiC, and ZrC comprises the coating materials section. Thermal, mechanical and physical properties of these materials were collected, reviewed, and summarized. Additionally, the property change induced by manufacture process and irradiation were collected and summarized. Performance data were also collected, reviewed, and analyzed based on the key phenomena and failure mechanism. All of these data will be accessible in the on-line system. These results will be directly used for HTGR fuel design and fabrication and preliminary fuel performance analysis under irradiation

  13. Irradiation effects on thermal properties of LWR hydride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  14. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  15. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    International Nuclear Information System (INIS)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs

  16. Determination of Basic Structure-Property Relations for Processing and Modeling in Advanced Nuclear Fuel: Microstructure Evolution and Mechanical Properties

    International Nuclear Information System (INIS)

    Wheeler, Kirk; Parra, Manuel; Peralta, Pedro

    2009-01-01

    The project objective is to study structure-property relations in solid solutions of nitrides and oxides with surrogate elements to simulate the behavior of fuels of inert matrix fuels of interest to the Advanced Fuel Cycle Initiative (AFCI), with emphasis in zirconium-based materials. Work with actual fuels will be carried out in parallel in collaboration with Los Alamos National Laboratory (LANL). Three key aspects will be explored: microstructure characterization through measurement of global texture evolution and local crystallographic variations using Electron Backscattering Diffraction (EBSD); determination of mechanical properties, including fracture toughness, quasi-static compression strength, and hardness, as functions of load and temperature, and, finally, development of structure-property relations to describe mechanical behavior of the fuels based on experimental data. Materials tested will be characterized to identify the mechanisms of deformation and fracture and their relationship to microstructure and its evolution. New aspects of this research are the inclusion of crystallographic information into the evaluation of fuel performance and the incorporation of statistical variations of microstructural variables into simplified models of mechanical behavior of fuels that account explicitly for these variations. The work is expected to provide insight into processing conditions leading to better fuel performance and structural reliability during manufacturing and service, as well as providing a simplified testing model for future fuel production

  17. Effect of flexible fuels on mechanical properties of reinforced polyoxymethylenes (POM

    Directory of Open Access Journals (Sweden)

    M. Gómez-Mares

    2014-08-01

    Full Text Available The use of flexible fuels has been increased during the last years making essential to run compatibility tests with those materials exposed to them. In this work the effect of the flexible fuels M15A (Volume Mixture of 85% fuel C and 15 % Aggressive methanol and M30A (Volume mixture of 70% fuel C and 30 % Aggressive methanol on the mechanical properties of some polymers of the Polyoxymethylene (POM family is assessed. The polymers chosen had different levels of glass fiber filler (0, 10 and 25%. The samples were immersed on fuel and kept on a chamber at 80°C during 1008h. The results showed that the properties of polymers with filler are more affected than the ones of the polymers without it. Tensile stress at break and Tensile stress at yield diminished with the fuel exposure. The most aggressive fuel was found to be M30A, due to the higher methanol concentration.

  18. Development of a process for co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder using microwave heating method

    International Nuclear Information System (INIS)

    Koizumi, Masumichi; Ohtsuka, Katsuyuki; Ohshima, Hirofumi; Isagawa, Hiroto; Akiyama, Hideo; Todokoro, Akio; Naruki, Kaoru

    1983-01-01

    For the complete nuclear fuel cycle, the development of a process for the co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder has been performed along the line of non-proliferation policy of nuclear materials. A new co-conversion process using a microwave heating method has been developed and successfully demonstrated with good results using the test unit with a capacity of 2 kg MOX/d. Through the experiments and engineering test operations, several important data have been obtained concerning the feasibility of the test unit, powder characteristics and homogeneity of the product, and impurity pickups during denitration process. The results of these experimental operations show that the co-conversion process using a microwave heating method has many excellent advantages, such as good powder characteristics of the product, good homogeneity of Pu-U oxide, simplicity of the process, minimum liquid waste, no possibility of changing the Pu/U ratio and stable operability of the plant. Since August 1979, plutonium nitrate solution transported from the Tokai Reprocessing Plant has been converted to mixed oxide powder which has the Pu/U ratio = 1. The products have been processed to the ATR ''FUGEN'' reloading fuel. Based on the successful development of the co-conversion process, the microwave heating direct denitration facility with a 10 kg MOX/d capacity has been constructed adjacent to the reprocessing plant. This facility will come into hot operation by the fall of this year. For future development of the microwave heating method, a continuous direct denitration, a vitrification of high active liquid waste and a solidification of the plutonium-contaminated waste are investigated in Power Reactor and Nuclear Fuel Development Corp. (author)

  19. TRISO coated fuel particles with enhanced SiC properties

    International Nuclear Information System (INIS)

    Lopez-Honorato, E.; Tan, J.; Meadows, P.J.; Marsh, G.; Xiao, P.

    2009-01-01

    The silicon carbide (SiC) layer used for the formation of TRISO coated fuel particles is normally produced at 1500-1650 deg. C via fluidized bed chemical vapor deposition from methyltrichlorosilane in a hydrogen environment. In this work, we show the deposition of SiC coatings with uniform grain size throughout the coating thickness, as opposed to standard coatings which have larger grain sizes in the outer sections of the coating. Furthermore, the use of argon as the fluidizing gas and propylene as a carbon precursor, in addition to hydrogen and methyltrichlorosilane, allowed the deposition of stoichiometric SiC coatings with refined microstructure at 1400 and 1300 deg. C. The deposition of SiC at lower deposition temperatures was also advantageous since the reduced heat treatment was not detrimental to the properties of the inner pyrolytic carbon which generally occurs when SiC is deposited at 1500 deg. C. The use of a chemical vapor deposition coater with four spouts allowed the deposition of uniform and spherical coatings.

  20. Durability Properties of Palm Oil Fuel Ash Self Compacting Concrete

    Directory of Open Access Journals (Sweden)

    T. Ofuyatan

    2015-02-01

    Full Text Available Self Compacting Concrete (SCC is a new innovation in technology that can flow readily into place under its own self weight and fill corner areas of reinforcement structures without the need to vibrate and without segregation of its constitute. The problem of durability of concrete structures due to inadequate compaction by skilled workers has become a source of concern globally. The shortage of skilled manpower, noise and vibration of equipment on construction sites has led to the development of self compacting concrete. This paper presents an experimental study on the durability properties of Self Compacting Concrete with partial placement of Palm Oil Fuel Ash (POFA. Twelve POFA self-compacting concretes of various strength grades were designed at varying percentages of 0, 5, 10, 15, 20, 25 and 30%. The concrete with no placement of ash served as control. Conplast SP432MS was used as superplasticiser in the mix. The experiments are carried out by adopting a water-powder ratio of 0.36. Workability of the fresh concrete is determined by using tests such as: slump flow, T50, V-funnel and L-Box tests. The durability of concrete is tested by acid resistance, sulphate attack and saturated water absorption at the age of 14, 28, 56 and 90 days.

  1. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  2. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  3. Isotopic composition and radiological properties of uranium in selected fuel cycles

    International Nuclear Information System (INIS)

    Fleischman, R.M.; Liikala, R.C.

    1975-04-01

    Three major topic areas are discussed: First, the properties of the uranium isotopes are defined relative to their respective roles in the nuclear fuel cycle. Secondly, the most predominant fuel cycles expected in the U. S. are described. These are the Light Water Reactor (LWR), High Temperature Gas Cooled Reactor (HTGR), and Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles. The isotopic compositions of uranium and plutonium fuels expected for these fuel cycles are given in some detail. Finally the various waste streams from these fuel cycles are discussed in terms of their relative toxicity. Emphasis is given to the high level waste streams from reprocessing of spent fuel. Wastes from the various fuel cycles are compared based on projected growth patterns for nuclear power and its various components. (U.S.)

  4. A Review of Hydrothermal Liquefaction Bio-Crude Properties and Prospects for Upgrading to Transportation Fuels

    Directory of Open Access Journals (Sweden)

    Jerome A. Ramirez

    2015-07-01

    Full Text Available Hydrothermal liquefaction (HTL presents a viable route for converting a vast range of materials into liquid fuel, without the need for pre-drying. Currently, HTL studies produce bio-crude with properties that fall short of diesel or biodiesel standards. Upgrading bio-crude improves the physical and chemical properties to produce a fuel corresponding to diesel or biodiesel. Properties such as viscosity, density, heating value, oxygen, nitrogen and sulphur content, and chemical composition can be modified towards meeting fuel standards using strategies such as solvent extraction, distillation, hydrodeoxygenation and catalytic cracking. This article presents a review of the upgrading technologies available, and how they might be used to make HTL bio-crude into a transportation fuel that meets current fuel property standards.

  5. Heat capacity measurements and XPS studies on uranium-lanthanum mixed oxides

    International Nuclear Information System (INIS)

    Venkata Krishnan, R.; Mittal, V.K.; Babu, R.; Senapati, Abhiram; Bera, Santanu; Nagarajan, K.

    2011-01-01

    Research highlights: → Heat capacity measurements were carried out on (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298-800 K. → Enthalpy increment measurements were carried out on the above solid solution using high temperature drop calorimetry in the temperature range 800-1800 K. → Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). → The anomalous increase in the heat capacity is attributed to certain thermal excitation process namely Frenkel pair defect of oxygen. → From the XPS investigation, it is observed that the O/M ratio at the surface is higher than that to the bulk. → In uranium rich mixed oxide samples, the surface O/M is greater than 2 whereas that in La rich mixed oxides, it is less than 2, though the bulk O/M in all the samples are less than 2. - Abstract: Heat capacity measurements were carried out on (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) using differential scanning calorimeter (DSC) in the temperature range 298-800 K. Enthalpy increment measurements were carried out on the above solid solutions using high temperature drop calorimetry in the temperature range 800-1800 K. Chemical states of U and La in the solid solutions of mixed oxides were determined using X-ray photoelectron spectroscopy (XPS). Oxygen to metal ratios of (U 1-y La y )O 2±x were estimated from the ratios of different chemical states of U present in the sample. Anomalous increase in the heat capacity is observed for (U 1-y La y )O 2±x (y = 0.4, 0.6 and 0.8) with onset temperatures in the range of 1000-1200 K. The anomalous increase in the heat capacity is attributed to certain thermal excitation process, namely, Frenkel pair defect of oxygen. The heat capacity value of (U 1-y La y )O 2±x (y = 0.2, 0.4, 0.6, and 0.8) at 298 K are 65.3, 64.1, 57.7, 51.9 J K -1 mol -1 , respectively. From the XPS investigations

  6. System for assembling nuclear fuel elements

    International Nuclear Information System (INIS)

    1980-01-01

    An automatic system is described for assembling nuclear fuel elements, in particular those employing mixed oxide fuels. The system includes a sealing mechanism which allows movement during the assembling of the fuel element along the assembly stations without excessive release of contaminants. (U.K.)

  7. Quality assurance for breeder reactor fuel fabrication

    International Nuclear Information System (INIS)

    Marx, E.R.

    1978-01-01

    Fuel performance in the Fast Flux Test Facility (FFTF) depends on fabrication of fuel to rigorous quality standards. The quality program including Management, Procurement, Fabrication, Inspection, Records, and Audits is discussed as well as unique mixed oxide fuel inspections such as homogeneity inspection, analytical chemistry, and nondestructive fissile assay

  8. A manufacturing process for a mixed-oxide type superconducting material

    International Nuclear Information System (INIS)

    Gendre, P.; Regnier, P.; Schmirgeld-Mignot, L.; Marquet, A.

    1995-01-01

    In order to produce high temperature superconducting materials such as YBaCuO and Bi 2 Sr 2 Ca Cu 2 O 8 , a process is presented which consists in an electrodeposition on a conductive substrate of successive layers made of the metallic elements composing the superconductor, with only one element in each layer; between each layer deposition, an intermediary oxide-reaction thermal treatment is carried out; a global oxidation thermal treatment is then finally conducted to produce the mixed oxide material. Narrow superconducting transitions and high critical current densities are possible. 3 refs., 4 figs

  9. Supported Mixed Oxide Catalysts for the Total Oxidation of Volatile Organic Compounds

    Czech Academy of Sciences Publication Activity Database

    Kovanda, F.; Jirátová, Květa

    2011-01-01

    Roč. 176, č. 1 (2011), s. 110-115 ISSN 0920-5861. [International Symposium on Air Pollution Abatement Catalysis (APAC) /2./. Cracow, 08.09.2010-10.09.2010] R&D Projects: GA ČR GAP106/10/1762; GA ČR GA106/09/1664 Institutional research plan: CEZ:AV0Z40720504 Keywords : layered double hydroxides * mixed oxides * ethanol total oxidation Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 3.407, year: 2011

  10. Investigation of Lubrication Properties of Petroleum Fuel and Biohydrocarbon Blends

    Directory of Open Access Journals (Sweden)

    Gawron Bartosz

    2016-07-01

    Full Text Available The paper covers issues regarding lubricity of petroleum fuels used in piston and turbine engines, containing hydrocarbon biocomponents. Basing on available literature it can be said that the most prospective fuel components are biohydrocarbons. The paper describes effect of biohydrocarbons included in aviation fuel and diesel fuel on lubricity of such blends. The analysis covers two processes for obtaining biohydrocarbons, the HVO and the Fischer-Tropsch process. Due to problems with actual products acquiring, biohydrocarbons models representing chemically the actual ones from specific process. Lubricity testing was carried out according to standard test methods.

  11. Expedient Prediction of the Fuel Properties of Carbonized Woody Biomass Based on Hue Angle

    Directory of Open Access Journals (Sweden)

    Yuta Saito

    2018-05-01

    Full Text Available Woody biomass co-firing-based power generation can reduce CO2 emissions from pulverized coal boilers. Carbonization of woody biomass increases its calorific value and grindability, thereby improving the co-firing ratio. Carbonized biomass fuel properties depend on moisture, size and shape of feedstock, and carbonization conditions. To produce carbonized biomass with stable fuel properties, the carbonization conditions should be set according to the desired fuel properties. Therefore, we examined color changes accompanying woody biomass carbonization and proposed using them for rapid evaluation of fuel properties. Three types of woody biomasses were carbonized at a test facility with a capacity of 4 tons/day, and the fuel properties of the obtained materials were correlated with their color defined by the L*a*b* model. When fixed carbon, an important fuel property for carbonization, was 25 wt % or less, we observed a strong negative correlation, regardless of the tree species, between the hue angle, hab, and fixed carbon. The hab and fixed carbon were correlated even when the fixed carbon exceeded 25 wt %; however, this correlation was specific to the tree species. These results indicate that carbonized biomass fuel properties such as fixed carbon can be estimated rapidly and easily by measuring hab.

  12. A study on properties-performances of coated particle fuel and on-line DB establishment

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Hyo Cheol; Jang, Jeong Nam; Kwon, Seok Hwan

    2007-03-01

    Recently national project for HTGR for hydrogen production has been kicked off. However, For the successful development of the high temperature gas cooled reactor high temperature and burn-up dependent properties of the reactor materials are essentially and crucially required. Therefore, it was proposed to build up the materials properties and fuel performance data base. In this study, a phase - 1 properties and performance DB for coated particle fuel was developed. This database report consists two sections: materials properties and fuel performance. The materials properties has three parts: kernel materials, carbide coating materials, and fuel elements and graphite matrix. UO2 and UCO belong to kernel materials while PyC, SiC, and ZrC comprises the coating materials section. Thermal, mechanical and physical properties data of these materials were collected, reviewed, and summarized. Additionally, the property change induced by manufacture process and irradiation were reviewed. Fuel performance data were also collected, reviewed, and analyzed based on the key phenomena and failure mechanism, These performance data are divided into two: normal and accident. All of these data will be accessible in the pc based stand-alone system. These results will be directly used for HTGR fuel design and fabrication and preliminary fuel performance analysis under irradiation

  13. Physical Properties of Biomass Fuel Briquette from Oil Palm ...

    African Journals Online (AJOL)

    ADOWIE PERE

    of fuel briquettes in this study in order to supplement the energy mix of the nation. PKS was pulverized ... fossil fuel in the world market is impacting negatively ... useful products that can be applied in many sectors ... at 350 µm, 250 µm and 150 µm with Octagon digital ... formula is one of the models developed to accurately.

  14. Fuel properties effect on the performance of a small high temperature rise combustor

    Science.gov (United States)

    Acosta, Waldo A.; Beckel, Stephen A.

    1989-01-01

    The performance of an advanced small high temperature rise combustor was experimentally determined at NASA-Lewis. The combustor was designed to meet the requirements of advanced high temperature, high pressure ratio turboshaft engines. The combustor featured an advanced fuel injector and an advanced segmented liner design. The full size combustor was evaluated at power conditions ranging from idle to maximum power. The effect of broad fuel properties was studied by evaluating the combustor with three different fuels. The fuels used were JP-5, a blend of Diesel Fuel Marine/Home Heating Oil, and a blend of Suntec C/Home Heating Oil. The fuel properties effect on the performance of the combustion in terms of pattern factor, liner temperatures, and exhaust emissions are documented.

  15. Impact of receipt of coprocessed uranium/plutonium on advanced accountability concepts and fabrication facilities. Addendum 1 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Randall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to assess the effect of coprocessed UO 2 --PuO 2 feed on the observations made in the original Phase I effort and on the proposed Phase II program. The retention of plutonium mixed with uranium throughout the process was also considered. This addendum reports that coprocessed feed would have minimal effect on the DYMAC program, except in the areas of material specifications, starting material delivery schedule, and labor requirements. Each of these areas is addressed, as are the impact of coprocessed feed at a large fuel fabrication facility and the changes needed in the dirty scrap recovery process to maintain the lower plutonium levels which may be required by future nonproliferation philosophy. An amended schedule for Phase II is included

  16. Discrimination of irradiated MOX fuel from UOX fuel by multivariate statistical analysis of simulated activities of gamma-emitting isotopes

    Science.gov (United States)

    Åberg Lindell, M.; Andersson, P.; Grape, S.; Hellesen, C.; Håkansson, A.; Thulin, M.

    2018-03-01

    This paper investigates how concentrations of certain fission products and their related gamma-ray emissions can be used to discriminate between uranium oxide (UOX) and mixed oxide (MOX) type fuel. Discrimination of irradiated MOX fuel from irradiated UOX fuel is important in nuclear facilities and for transport of nuclear fuel, for purposes of both criticality safety and nuclear safeguards. Although facility operators keep records on the identity and properties of each fuel, tools for nuclear safeguards inspectors that enable independent verification of the fuel are critical in the recovery of continuity of knowledge, should it be lost. A discrimination methodology for classification of UOX and MOX fuel, based on passive gamma-ray spectroscopy data and multivariate analysis methods, is presented. Nuclear fuels and their gamma-ray emissions were simulated in the Monte Carlo code Serpent, and the resulting data was used as input to train seven different multivariate classification techniques. The trained classifiers were subsequently implemented and evaluated with respect to their capabilities to correctly predict the classes of unknown fuel items. The best results concerning successful discrimination of UOX and MOX-fuel were acquired when using non-linear classification techniques, such as the k nearest neighbors method and the Gaussian kernel support vector machine. For fuel with cooling times up to 20 years, when it is considered that gamma-rays from the isotope 134Cs can still be efficiently measured, success rates of 100% were obtained. A sensitivity analysis indicated that these methods were also robust.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    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

  18. Alternative fuel properties of tall oil fatty acid methyl ester-diesel fuel blends

    Energy Technology Data Exchange (ETDEWEB)

    Altiparmak, D.; Keskin, A.; Koca, A. [Gazi University, Ankara (Turkey). Technical Education Faculty; Guru, M. [Gazi University, Ankara (Turkey). Engineering and Architectural Faculty

    2007-01-15

    In this experimental work, tall oil methyl ester-diesel fuel blends as alternative fuels for diesel engines were studied. Tall oil methyl ester was produced by reacting tall oil fatty acids with methyl alcohol under optimum conditions. The blends of tall oil methyl ester-diesel fuel were tested in a direct injection diesel engine at full load conditions. The effects of the new fuel blends on the engine performance and exhaust emission were tested. It was observed that the engine torque and power output with tall oil methyl ester-diesel fuel blends increased up to 6.1% and 5.9%, respectively. It was also seen that CO emissions decreased to 38.9% and NO{sub x} emissions increased up to 30% with the new fuel blends. The smoke capacity did not vary significantly. (author)

  19. Alternative fuel properties of tall oil fatty acid methyl ester-diesel fuel blends.

    Science.gov (United States)

    Altiparmak, Duran; Keskin, Ali; Koca, Atilla; Gürü, Metin

    2007-01-01

    In this experimental work, tall oil methyl ester-diesel fuel blends as alternative fuels for diesel engines were studied. Tall oil methyl ester was produced by reacting tall oil fatty acids with methyl alcohol under optimum conditions. The blends of tall oil methyl ester-diesel fuel were tested in a direct injection diesel engine at full load condition. The effects of the new fuel blends on the engine performance and exhaust emission were tested. It was observed that the engine torque and power output with tall oil methyl ester-diesel fuel blends increased up to 6.1% and 5.9%, respectively. It was also seen that CO emissions decreased to 38.9% and NO(x) emissions increased up to 30% with the new fuel blends. The smoke opacity did not vary significantly.

  20. Biodiesel production from various feedstocks and their effects on the fuel properties.

    Science.gov (United States)

    Canakci, M; Sanli, H

    2008-05-01

    Biodiesel, which is a new, renewable and biological origin alternative diesel fuel, has been receiving more attention all over the world due to the energy needs and environmental consciousness. Biodiesel is usually produced from food-grade vegetable oils using transesterification process. Using food-grade vegetable oils is not economically feasible since they are more expensive than diesel fuel. Therefore, it is said that the main obstacle for commercialization of biodiesel is its high cost. Waste cooking oils, restaurant greases, soapstocks and animal fats are potential feedstocks for biodiesel production to lower the cost of biodiesel. However, to produce fuel-grade biodiesel, the characteristics of feedstock are very important during the initial research and production stage since the fuel properties mainly depend on the feedstock properties. This review paper presents both biodiesel productions from various feedstocks and their effects on the fuel properties.

  1. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  2. Ignition Delay Properties of Alternative Fuels with Navy-Relevant Diesel Injectors

    Science.gov (United States)

    2014-06-01

    nozzle tip. 8 Figure 3 EMD injector cross-sectional view, after [15]. c. Sturman Injector A Sturman research diesel injector was used to validate...PROPERTIES OF ALTERNATIVE FUELS WITH NAVY-RELEVANT DIESEL INJECTORS by Andrew J. Rydalch June 2014 Thesis Advisor: Christopher M. Brophy...Navy’s Green Fleet Initiative, this thesis researched the ignition characteristics for diesel replacement fuels used with Navy-relevant fuel injectors

  3. Ytterbia doped nickel–manganese mixed oxide catalysts for liquid phase oxidation of benzyl alcohol

    Directory of Open Access Journals (Sweden)

    S.S.P. Sultana

    2017-11-01

    Full Text Available Nickel–manganese mixed oxides doped with 1, 3, 5 mol% ytterbia have been prepared by co-precipitation method and used in the catalytic oxidation of benzyl alcohol. Catalytic activity of these oxides calcined at 400 °C and 500 °C was studied for selective oxidation of benzyl alcohol to the corresponding aldehyde using molecular oxygen as an oxidizing agent. The results showed that thermally stable 5 mol% ytterbia doped nickel–manganese oxide [Yb2O3-(5%-Ni6MnO8] exhibited highest catalytic performance when it was calcined at 400 °C. A 100% conversion of the benzyl alcohol was achieved with >99% selectivity to benzaldehyde within a reaction period of 5 h at 100 °C. The mixed oxide prepared has been characterized by scanning election microscopy (SEM and energy dispersive X-ray analysis (EDXA, X-ray diffraction (XRD, Fourier transform infrared spectroscopy (FT-IR, thermogravimetric analysis (TGA, Brunauer–Emmett–Teller (BET and temperature programed reduction (H2-TPR.

  4. Impact of future fuel properties on aircraft engines and fuel systems

    Science.gov (United States)

    Rudey, R. A.; Grobman, J. S.

    1978-01-01

    From current projections of the availability of high-quality petroleum crude oils, it is becoming increasingly apparent that the specifications for hydrocarbon jet fuels may have to be modified. The problems that are most likely to be encountered as a result of these modifications relate to engine performance, component durability and maintenance, and aircraft fuel-system performance. The effect on engine performance will be associated with changes in specific fuel consumption, ignition at relight limits, at exhaust emissions. Durability and maintenance will be affected by increases in combustor liner temperatures, carbon deposition, gum formation in fuel nozzles, and erosion and corrosion of turbine blades and vanes. Aircraft fuel-system performance will be affected by increased deposits in fuel-system heat exchangers and changes in the pumpability and flowability of the fuel. The severity of the potential problems is described in terms of the fuel characteristics most likely to change in the future. Recent data that evaluate the ability of current-technology aircraft to accept fuel specification changes are presented, and selected technological advances that can reduce the severity of the problems are described and discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-27

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-14

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

  7. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  8. Nuclear-fuel-cycle education: Module 10. Environmental consideration

    International Nuclear Information System (INIS)

    Wethington, J.A.; Razvi, J.; Grier, C.; Myrick, T.

    1981-12-01

    This educational module is devoted to the environmental considerations of the nuclear fuel cycle. Eight chapters cover: National Environmental Policy Act; environmental impact statements; environmental survey of the uranium fuel cycle; the Barnwell Nuclear Fuel Reprocessing Plant; transport mechanisms; radiological hazards in uranium mining and milling operations; radiological hazards of uranium mill tailings; and the use of recycle plutonium in mixed oxide fuel

  9. Transesterification of Jatropha oil with dimethyl carbonate to produce fatty acid methyl ester over reusable Ca–La–Al mixed-oxide catalyst

    International Nuclear Information System (INIS)

    Syamsuddin, Y.; Murat, M.N.; Hameed, B.H.

    2015-01-01

    Highlights: • Transesterification of Jatropha oil over CaO-based catalyst. • Physicochemical properties of the synthesized catalyst. • Best reaction condition for FAME synthesis. • The catalyst showed high activity and stability for transesterification with Jatropha oil. - Abstract: Jatropha oil (JO) was transesterified with dimethyl carbonate (DMC) to produce fatty acid methyl ester (FAME) over synthesized Ca–La–Al mixed-oxide catalyst. The influence of different parameters on transesterification of Jatropha oil was investigated in a batch reactor. These parameters included reaction temperature (110–160 °C), reaction time (30–240 min), DMC-to-oil molar ratio (4:1–18:1) and catalyst loading amount (1–10 wt.%, based on the oil weight). The mixed-oxide catalyst with a molar ratio of 6:2:1 (Ca–La–Al) showed high catalytic activity for FAME synthesis. More than 90% of FAME was obtained under the following reaction conditions: 150 °C, reaction temperature; 180 min, reaction time; 15:1, DMC-to-oil molar ratio; and 7 wt.% amount of catalyst loading. The catalyst also exhibited high stability and could be reused for up to five cycles with less than 5% yield reduction per cycle.

  10. Non-destructive methods of control of thermo-physical properties of fuel rods

    International Nuclear Information System (INIS)

    Kruglov, A B; Kruglov, V B; Kharitonov, V S; Struchalin, P G; Galkin, A G

    2017-01-01

    Information about the change of thermal properties of the fuel elements needed for a successful and safe operation of the nuclear power plant. At present, the existing amount of information on the fuel thermal conductivity change and “fuel-shell” thermal resistance is insufficient. Also, there is no technique that would allow for the measurement of these properties on the non-destructive way of irradiated fuel elements. We propose a method of measuring the thermal conductivity of the fuel in the fuel element and the contact thermal resistance between the fuel and the shell without damaging the integrity of the fuel element, which is based on laser flash method. The description of the experimental setup, implementing methodology, experiments scheme. The results of test experiments on mock-ups of the fuel elements and their comparison with reference data, as well as the results of numerical modeling of thermal processes that occur during the measurement. Displaying harmonization of numerical calculation with the experimental thermograms layout shell portions of the fuel cell, confirming the correctness of the calculation model. (paper)

  11. Prediction of air-fuel and oxy-fuel combustion through a generic gas radiation property model

    International Nuclear Information System (INIS)

    Yin, Chungen

    2017-01-01

    Highlights: • A gas radiation model for general combustion CFD presented, programmed & verified. • Its general applicability/practical accuracy demonstrated in air-fuel and oxy-fuel. • Useful guidelines for air-fuel and oxy-fuel combustion CFD suggested. • Important to include the impact of CO in gas radiation for oxy-fuel combustion CFD. - Abstract: Thermal radiation plays an important role in heat transfer in combustion furnaces. The weighted-sum-of-gray-gases model (WSGGM), representing a good compromise between computational efficiency and accuracy, is commonly used in computational fluid dynamics (CFD) modeling of combustion processes for evaluating gaseous radiative properties. However, the WSGGMs still have some limitations in practical use, e.g., unable to naturally accommodate different combustion environments, difficult to accurately address the variations in species concentrations in a flame, and inconvenient to account for the impacts of participating species other than H_2O and CO_2. As a result, WSGGMs with different coefficients have been published for specific applications. In this paper, a reliable generic model for gaseous radiation property calculation, which is a computationally efficient exponential wide band model (E-EWBM) applicable to combustion CFD and able to naturally solve all the practical limitations of the WSGGMs, is presented, programmed and verified. The model is then implemented to CFD simulation of a 300 kW air-fuel and a 0.8 MW oxy-fuel combustion furnace, respectively, to demonstrate its computational applicability to general combustion CFD and its capability in producing reliable CFD results for different combustion environments. It is found that the usefulness of the WSGGMs in oxy-fuel combustion CFD is compromised if the important impacts of high levels of CO under oxy-fuel combustion cannot be accounted for. The E-EWBM that appropriately takes the impacts of H_2O, CO_2, CO and CH_4 into account is a good replacement

  12. Sensitivities of Internal Combustion Automotive Engines to Variations in Fuel Properties

    Science.gov (United States)

    1982-02-01

    An assessment of the sensitivity of the automotive gasoline and diesel engines to variations in fuel properties has been made. The variables studied include H/C ratio, distillation range, aromatic content, ignition quality as determined by the octane...

  13. The data-base of properties of actinides for metal fuels

    International Nuclear Information System (INIS)

    Inoue, Tadashi; Kurata, Masateru

    1989-01-01

    It is developed the technology that transuranium elements (TRUs) to be recovered from high active wastes transmute into relatively short lived nuclides by burning them within metallic fuel alloys. In this paper, we collect published data of properties of TRUs and U-Pu(-Zr) alloys and make up the data base for the design study of alloys with TRUs. In addition, the data base possesses a function of statistic analysis in order to facilitate the comparison of data and can afford to estimate properties. This data base collects (1) properties affecting fuel temperature and microstructure, (2) mechanical properties and (3) fundamental properties such as hardness and density, and furthermore, (1) fission gas release, (2) swelling and (3) fuel-cladding interaction and eutectic property as irradiation behavior. (author)

  14. Research in Supercritical Fuel Properties and Combustion Modeling

    Science.gov (United States)

    2015-09-18

    identified reactions needing further study and C-2 and C-3 species to add to the mechanism . 15. SUBJECT TERMS Supercritical fluids , Brillouin scattering...kinetics mechanism for combustion of hydrocarbon fuels containing up to 2 carbon atoms, including uncertainties. • We identified key reactions and...safety. The chemical mechanisms for combustion of all of these fuels share the same set of elementary reactions of smaller-fragment hydrocarbons , and

  15. A computational methodology for formulating gasoline surrogate fuels with accurate physical and chemical kinetic properties

    KAUST Repository

    Ahmed, Ahfaz

    2015-03-01

    Gasoline is the most widely used fuel for light duty automobile transportation, but its molecular complexity makes it intractable to experimentally and computationally study the fundamental combustion properties. Therefore, surrogate fuels with a simpler molecular composition that represent real fuel behavior in one or more aspects are needed to enable repeatable experimental and computational combustion investigations. This study presents a novel computational methodology for formulating surrogates for FACE (fuels for advanced combustion engines) gasolines A and C by combining regression modeling with physical and chemical kinetics simulations. The computational methodology integrates simulation tools executed across different software platforms. Initially, the palette of surrogate species and carbon types for the target fuels were determined from a detailed hydrocarbon analysis (DHA). A regression algorithm implemented in MATLAB was linked to REFPROP for simulation of distillation curves and calculation of physical properties of surrogate compositions. The MATLAB code generates a surrogate composition at each iteration, which is then used to automatically generate CHEMKIN input files that are submitted to homogeneous batch reactor simulations for prediction of research octane number (RON). The regression algorithm determines the optimal surrogate composition to match the fuel properties of FACE A and C gasoline, specifically hydrogen/carbon (H/C) ratio, density, distillation characteristics, carbon types, and RON. The optimal surrogate fuel compositions obtained using the present computational approach was compared to the real fuel properties, as well as with surrogate compositions available in the literature. Experiments were conducted within a Cooperative Fuels Research (CFR) engine operating under controlled autoignition (CAI) mode to compare the formulated surrogates against the real fuels. Carbon monoxide measurements indicated that the proposed surrogates

  16. Mox fuels recycling

    International Nuclear Information System (INIS)

    Gay, A.

    1998-01-01

    This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

  17. Tuning the acidity of niobia: Characterization and catalytic activity of Nb2O5–MeO2 (Me = Ti, Zr, Ce) mesoporous mixed oxides

    International Nuclear Information System (INIS)

    Stošić, Dušan; Bennici, Simona; Pavlović, Vladimir; Rakić, Vesna; Auroux, Aline

    2014-01-01

    Mesoporous Nb 2 O 5 –MeO 2 (Me = Ti, Zr, Ce) mixed oxides were successfully prepared using evaporation-induced self-assembly (EISA) method. The structural and textural properties of these materials have been fully characterized using appropriate techniques (low-temperature adsorption–desorption of nitrogen, thermogravimetric analysis, X-ray diffraction analysis (XRD) transmission electron microscopy (TEM), scanning electron microscopy (SEM) and Raman spectroscopy). Acid–base properties were estimated by adsorption microcalorimetry of NH 3 and SO 2 molecules in order to determine the population, strength and strength distribution of acidic or basic sites. Formation of mesoporous structure was confirmed by the results of XRD, TEM and BET techniques. Results of adsorption microcalorimetry technique showed that the type of transition metal oxide added to niobia has a decisive role for acidic-basic character of investigated mixed oxides. Among the investigated mixed oxide formulations only Nb 2 O 5 –CeO 2 was amphoteric, while the other samples showed prominent acidic character. All the investigated materials are catalytically active in fructose dehydration; conversion of fructose and selectivity to 5-hydroxymethylfurfural (5-HMF) and levulinic acid (LA) are proved to be dependant on the number of acidic sites on the surface of catalysts. Furthermore, presence of the basic sites on the surface of the catalyst decreases the activity in the fructose dehydration reaction, as in the case of Nb 2 O 5 –CeO 2 sample. - Highlights: • Mesoporous Nb 2 O 5 –MeO 2 mixed oxides were successfully prepared by EISA method. • Acidic–basic properties depend on the nature of the oxide that was mixed with niobia. • Catalytic activity was tested in fructose dehydration in aqueous phase. • Selectivity and conversion in reaction are correlated to the number of acid sites

  18. Mixed oxides obtained from Co and Mn containing layered double hydroxides: Preparation, characterization, and catalytic properties

    Czech Academy of Sciences Publication Activity Database

    Kovanda, F.; Rojka, T.; Dobešová, J.; Machovič, V.; Bezdička, Petr; Obalová, L.; Jirátová, Květa; Grygar, Tomáš

    2006-01-01

    Roč. 179, č. 3 (2006), s. 812-823 ISSN 0022-4596 R&D Projects: GA ČR(CZ) GA104/04/2116; GA ČR(CZ) GA106/05/0366 Institutional research plan: CEZ:AV0Z40320502; CEZ:AV0Z40720504 Keywords : layered double hydroxides * hydrotalcite-like compounds * thermal decomposition Subject RIV: CA - Inorganic Chemistry Impact factor: 2.107, year: 2006

  19. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    1988-06-01

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  20. Corrosion effect of fast reactor fuel claddings on their mechanical properties

    International Nuclear Information System (INIS)

    Davydov, E.F.; Krykov, F.N.; Shamardin, V.K.

    1985-01-01

    Fast reactor fuel cladding corrosion effect on its mechanical properties was investigated. UO 2 fuel elements were irradiated in the BOP-60 reactor at the linear heat rate of 42 kw/m. Fuel cladding is made of stainless steel OKh16N15M3BR. Calculated maximum cladding temperature is 920 K. Neutron fluence in the central part of fuel elements is 6.3x10 26 m+H- 2 . To investigate the strength changes temperature dependence of corrossion depth, cladding strength reduction factors was determined. Samples plasticity reduction with corrosion layer increase is considered to be a characteristic feature

  1. Influence of vanadium oxidation states on the performance of V-Mg-Al mixed-oxide catalysts for the oxidative dehydrogenation of propane

    International Nuclear Information System (INIS)

    Schacht, L.; Navarrete, J.; Schacht, P.; Ramirez, M. A.

    2010-01-01

    V-Mg-Al mixed-oxide catalysts for oxidative dehydrogenation of propane were prepared by thermal decomposition of Mg-Al-layered double hydroxides with vanadium interlayer doping. The obtained catalysts were tested for the oxidative dehydrogenation of propane, obtaining good results in catalytic activity (conversion 16.55 % and selectivity 99.97 %) Results indicated that catalytic performance of these materials depends on how vanadium is integrated in the layered structure, which is determined by the Mg/Al ratio. Vanadium interlayer doping modifies the oxidation state of vanadium and consequently catalytic properties. Surface properties were studied by X-ray photoelectron spectroscopic and diffuse reflectance, UV-visible spectroscopy, and temperature programmed reduction. The analyses provided information about the oxidation state, before and after the reaction. From these results, it is suggested that selectivity to propylene and catalytic activity depend mainly of vanadium oxidation state. (Author)

  2. Influence of vanadium oxidation states on the performance of V-Mg-Al mixed-oxide catalysts for the oxidative dehydrogenation of propane

    Energy Technology Data Exchange (ETDEWEB)

    Schacht, L. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ciencia de Materiales, Av. IPN s/n, Edificio 9, Col. Lindavista, 07738 Mexico D. F. (Mexico); Navarrete, J.; Schacht, P.; Ramirez, M. A., E-mail: pschacha@imp.m [Instituto Mexicano del Petroleo, Programa de Ingenieria Molecular, Eje Central Lazaro Cardenas No. 152, 07730 Mexico D. F. (Mexico)

    2010-07-01

    V-Mg-Al mixed-oxide catalysts for oxidative dehydrogenation of propane were prepared by thermal decomposition of Mg-Al-layered double hydroxides with vanadium interlayer doping. The obtained catalysts were tested for the oxidative dehydrogenation of propane, obtaining good results in catalytic activity (conversion 16.55 % and selectivity 99.97 %) Results indicated that catalytic performance of these materials depends on how vanadium is integrated in the layered structure, which is determined by the Mg/Al ratio. Vanadium interlayer doping modifies the oxidation state of vanadium and consequently catalytic properties. Surface properties were studied by X-ray photoelectron spectroscopic and diffuse reflectance, UV-visible spectroscopy, and temperature programmed reduction. The analyses provided information about the oxidation state, before and after the reaction. From these results, it is suggested that selectivity to propylene and catalytic activity depend mainly of vanadium oxidation state. (Author)

  3. Airbreathing Propulsion Fuels and Energy Exploratory Research and Development (APFEERD) Sub Task: Review of Bulk Physical Properties of Synthesized Hydrocarbon:Kerosenes and Blends

    Science.gov (United States)

    2017-06-01

    Fuels and Energy Branch Turbine Engine Division Turbine Engine Division CHARLES W. STEVENS, Lead Engineer Turbine Engine Division Aerospace Systems...evaluation concludes, based on fundamental physical chemistry , that all hydrocarbon kerosenes that meet the minimum density requirement will have bulk...alternative jet fuels; renewable jet fuel; fuel physical properties; fuel chemistry ; fuel properties 16. SECURITY CLASSIFICATION OF: 17. LIMITATION OF

  4. Thermal Properties of Green Fuel Briquettes from Residue Corncobs Materials Mixed Macadamia Shell Charcoal Powder

    Science.gov (United States)

    Teeta, Suminya; Nachaisin, Mali; Wanish, Suchana

    2017-09-01

    The objective of this research was to produce green fuel briquettes from corncobs by adding macadamia shell charcoal powder. The study was sectioned into 3 parts: 1) Quality improvement of green fuel briquettes by adding macadamia; 2) Fuel property analysis based on ASTM standards and thermal fuel efficiency; and 3) Economics appropriateness in producing green fuel briquettes. This research produced green fuel briquettes using the ratio of corncobs weight and macadamia shell charcoal powder in 100:0 90:10 80:20 70:30 60:40 and 50:50 and pressing in the cold briquette machine. Fuel property analysis showed that green fuel briquettes at the ratio 50:50 produced maximum heating values at 21.06 Megajoule per kilogram and briquette density of 725.18 kilograms per cubic meter, but the percent of moisture content, volatile matter, ash, and fixed carbon were 10.09, 83.02, 2.17 and 4.72 respectively. The thermal efficiency of green fuel briquettes averaged 20.22%. Economics appropriateness was most effective where the ratio of corncobs weight to macadamia shell charcoal powder was at 50:50 which accounted for the cost per kilogram at 5.75 Baht. The net present value was at 1,791.25 Baht. Internal rate of return was at 8.62 and durations for a payback period of investment was at 1.9 years which was suitable for investment.

  5. International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    Lovasic, Z.; Einziger, R.

    2009-01-01

    This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

  6. The fractalline properties of experimentally simulated PWR fuel crud

    Science.gov (United States)

    Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

    2018-02-01

    The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

  7. Storage and characterization of the hydrogen in mixed oxides on base of cerium-nickel and zirconium or the aluminium

    International Nuclear Information System (INIS)

    Debeusscher, S.

    2008-12-01

    The mixed oxides based on cerium-nickel and zirconium or aluminium are able to store large quantities of hydrogen, To determine nature, reactivity and properties of hydrogen species (spill-over, direct desorption), the solid were studied by different physicochemical techniques in the dried, calcined and partially reduced states: XRD, porosity, TGA, TPR, TPA, TPD, chemical titration and inelastic neutron scattering (INS). Solids are mainly meso-porous with a common pore size at 4 nm, They are constituted of CeO 2 phase, Ce-Ni or Ce-Ni-Zr solid solution and of Ni(OH) 2 in the dried state and NiO in the calcined state. The Ni species are in various environments and the strong interactions between the cations in solid solution and at different particles interface influence their reducibility and the creation of anionic vacancies. Activation in H 2 in temperature is determining for hydrogen storage in the solid while calcination step is not necessary. INS Analyses evidence that the hydrogen species inserted during treatment in H 2 are H + (OH - ), hydride H - and H * (metallic nickel) species, present in various chemical environments, in particular for hydride species. All kinds of hydrogen species participate to the reaction during the chemical titration in agreement with the proposed hydrogenation mechanism. The study of the adsorption of hydrogen shows that this step is fast and in quantity of the same order as that measured by chemical titration. The direct desorption of H 2 is very low, linked to the presence of hydrogen in interaction with metallic nickel (H *- .). Desorption of water is also observed, in parallel, corresponding to the elimination of groups. The hydride species are not desorbed. These various observations allow connecting hydrogen species properties with their localization in the structure and to model active sites. (author)

  8. Influence of Gold on Ce-Zr-Co Fluorite-Type Mixed Oxide Catalysts for Ethanol Steam Reforming

    Directory of Open Access Journals (Sweden)

    Véronique Pitchon

    2012-02-01

    Full Text Available The effect of gold presence on carbon monoxide oxidation and ethanol steam reforming catalytic behavior of two Ce-Zr-Co mixed oxides catalysts with a constant Co charge and different Ce/Zr ratios was investigated. The Ce-Zr-Co mixed oxides were obtained by the pseudo sol-gel like method, based on metallic propionates polymerization and thermal decomposition, whereas the gold-supported Ce-Zr-Co mixed oxides catalysts were prepared using the direct anionic exchange. The catalysts were characterized using XRD, TPR, and EDXS-TEM. The presence of Au in doped Ce-Zr-Co oxide catalyst decreases the temperature necessary to reduce the cobalt and the cerium loaded in the catalyst and favors a different reaction pathway, improving the acetaldehyde route by ethanol dehydrogenation, instead of the ethylene route by ethanol dehydration or methane re-adsorption, thus increasing the catalytic activity and selectivity into hydrogen.

  9. Template-free synthesis of mesoporous nanoring-like Zn-Co mixed oxides with high lithium storage performance

    Science.gov (United States)

    Lu, Lun; Gao, Yan-Li; Yang, Zhi-Zheng; Wang, Cheng; Wang, Jin-Guo; Wang, Hui-Yuan; Jiang, Qi-Chuan

    2018-04-01

    Mesoporous nanoring-like Zn-Co mixed oxides are synthesized via a simple template-free solvothermal method with a subsequent annealing process. The ring-like nanostructures with hollow interiors are formed under the complexing effects of potassium sodium tartrate. Numerous mesopores are generated after the precursor is annealed at 500 °C. When applied as anode materials, the mesoporous nanoring-like Zn-Co mixed oxides can deliver a high discharge capacity of 1102 mAh g-1 after 200 cycles at 500 mA g-1. Even when the current density is increased to 2 A g-1, the mixed oxides can still retain a reversible capacity of 761 mAh g-1. Such high cycling stability and rate capability are mainly derived from the unique mesoporous ring-like nanostructures and the synergistic effects between Zn and Co based oxides.

  10. Design of the MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Johnson, J.V.; Brabazon, E.J.

    2001-01-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  11. Design of the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)

    2001-07-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  12. Determination of the oxygen-metal-ratio of uranium-americium mixed oxides

    International Nuclear Information System (INIS)

    Bartscher, W.

    1982-01-01

    During the dissolution of uranium-americium mixed oxides in phosphoric acid under nitrogen tetravalent uranium is oxidized by tetravalent americium. The obtained hexavalent uranium is determined by constant potential coulometry. The coulombs measured are equivalent to the oxygen in excess of the minimum composition of UO 2 x AmO 1 . 5 . The total uranium content of the sample is determined in a subsequent coulometric titration. The oxygen-metal ratio of the sample can be calculated for a given uranium-americium ratio. An excess of uranium dioxide is necessary in order to suppress the oxidation of water by tetravalent americium. The standard deviation of the method is 0.0017 O/M units. (orig.) [de

  13. Removal of inhaled industrial mixed oxide aerosols from Beagle dogs by lung lavage and chelation therapy

    International Nuclear Information System (INIS)

    Muggenburg, B.A.; Mewhinney, J.A.; Eidson, A.F.; Guilmette, R.A.

    1978-01-01

    An experiment was conducted in 15 adult Beagle dogs to evaluate lung lavage and chelation therapy for the removal of inhaled particles of mixed actinide oxides. The dogs were divided into three groups of five dogs each. Each group was exposed to an aerosol from a different industrial process. Group 1 was exposed to mixed oxide material which had been calcined at 750 0 C collected from a ball milling process. Group 2 was exposed to mixed oxide material from a centerless grinding operation which had been previously heat treated to 1750 0 C. The third group was exposed to 239 PuO 2 not containing uranium from a V-blending procedure which had been heat treated at 850 0 C. After exposure, three dogs in each group were given ten lung lavages and 18 intravenous injections of calcium trisodium diethylenetriaminepentaacetate (DTPA). All dogs were sacrificed 64 days after inhalation exposure. The tissues were radioanalyzed for plutonium and americium. Fluorimetric analyses for uranium in the tissues are in progress. The urine, feces and lavage fluid are also being analyzed for plutonium, americium and uranium. The distribution of plutonium and americium expressed as percentages of the sacrifice body burden was similar in the tissues of the treated and unteated dogs. The lungs contained most of the radionuclides with a small amount in the liver, skeleton and tracheobronchial lymph nodes. The percentage of the sacrifice body burden of americium and plutonium that was present in the lung was less in the treated dogs and was higher in the TBLN's and skeleton than in the untreated dogs. The ratio of Pu/Am was higher in the lungs than in the original material obtained from the industrial sites suggesting a shorter retention time for americium than plutonium to 64 days in the dog

  14. Ce-Fe-O mixed oxide as oxygen carrier for the direct partial oxidation of methane to syngas

    Institute of Scientific and Technical Information of China (English)

    魏永刚; 王华; 李孔斋

    2010-01-01

    The Ce-Fe-O mixed oxide with a ratio of Ce/Fe=7:3, which was prepared by coprecipitation method and employed as oxygen carrier, for direct partial oxidation of methane to syngas in the absence of gaseous oxygen was explored. The mixed oxide was characterized by X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS) and scanning electron microscopy (SEM), and the catalytic performances were studied in a fixed-bed quartz reactor and a thermogravimetric reactor, respectively. Approximately 99.4% H2 se...

  15. A comparative study on determination of composition of uranium thorium mixed oxides by tube and radioisotope excited EDXRF

    International Nuclear Information System (INIS)

    Dhara, Sangita; Sanjay Kumar, S.; Misra, N.L.; Aggarwal, S.K.; Singh, Ajit Kumar; Lodha, G.S.

    2009-01-01

    Energy Dispersive X-ray Fluorescence (EDXRF) methods for determination of uranium and thorium in their mixed oxide matrices using tube and radioisotope excitation sources have been developed. The methodology involves preparation of mixed oxide calibration/sample mixtures of uranium and thorium oxides, mixing of fixed amount of internal standard Yttrium in form of Yttrium oxide, pelletizing these mixtures after thorough mixing and recording their EDXRF spectra using Rh target as well as 109 Cd radioisotope source. The samples were analysed for uranium and thorium on the basis of calibration plots

  16. Properties of the high burnup structure in nuclear light water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wiss, Thierry; Rondinella, Vincenzo V.; Konings, Rudy J.M. [European Commission, Joint Research Centre, Karlsruhe (Germany). Directorate Nuclear Safety and Security; and others

    2017-07-01

    The formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2-3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

  17. Improved correlations of hydrogen content versus combustion performance related properties of aviation turbine fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nagpal, J.M.; Sharma, R.L.; Sagu, M.L.; Tiwari, G.B. (Indian Institute of Petroleum, Dehradun (India))

    1994-01-01

    In recent years the hydrogen content of Aviation Fuels has generated considerable interest. Various investigators have suggested correlation of hydrogen content with combustion related properties of aviation turbine fuel (ATF). A suitable threshold value of hydrogen content 13.8 wt% is being considered as a waiver of specifications such as specific energy, aniline gravity product, smoke point, aromatic content, naphthalenes and luminometer number. In the present paper relationship between the hydrogen content and combustion related properties has been examined and improved correlations of hydrogen content with several combustion related properties have been developed by incorporating a characterization factor in the equations. The supporting threshold value of a hydrogen content of 13.8wt% is verified with 25 data points for waiving of combustion properties such as specific energy, aniline gravity product, smoke point and aromatic content from aviation turbine fuel. 6 refs., 12 figs., 2 tabs.

  18. Influence of Chemical Blends on Palm Oil Methyl Esters’ Cold Flow Properties and Fuel Characteristics

    Directory of Open Access Journals (Sweden)

    Obed M. Ali

    2014-07-01

    Full Text Available Alternative fuels, like biodiesel, are being utilized as a renewable energy source and an effective substitute for the continuously depleting supply of mineral diesel as they have similar combustion characteristics. However, the use of pure biodiesel as a fuel for diesel engines is currently limited due to problems relating to fuel properties and its relatively poor cold flow characteristics. Therefore, the most acceptable option for improving the properties of biodiesel is the use of a fuel additive. In the present study, the properties of palm oil methyl esters with increasing additive content were investigated after addition of ethanol, butanol and diethyl ether. The results revealed varying improvement in acid value, density, viscosity, pour point and cloud point, accompanied by a slight decrease in energy content with an increasing additive ratio. The viscosity reductions at 5% additive were 12%, 7%, 16.5% for ethanol, butanol and diethyl ether, respectively, and the maximum reduction in pour point was 5 °C at 5% diethyl ether blend. Engine test results revealed a noticeable improvement in engine brake power and specific fuel consumption compared to palm oil biodiesel and the best performance was obtained with diethyl ether. All the biodiesel-additive blend samples meet the requirements of ASTM D6751 biodiesel fuel standards for the measured properties.

  19. Influence of plutonium contents in MOX fuel on destructive forces at fuel failure in the NSRR experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    In order to confirm safety margins of the Mixed Oxide (MOX) fuel use in LWRs, pulse irradiation tests are planned in the Nuclear Safety Research Reactor (NSRR) with the MOX fuel with plutonium content up to 12.8%. Impacts of the higher plutonium contents on safety of the reactivity-initiated-accident (RIA) tests are examined in terms of generation of destructive forces to threat the integrity of test capsules. Pressure pulses would be generated at fuel rod failure by releases of high pressure gases. The strength of the pressure pulses, therefore, depends on rod internal - external pressure difference, which is independent to plutonium content of the fuel. The other destructive forces, water hammer, would be generated by thermal interaction between fuel fragments and coolant water. Heat flux from the fragments to the water was calculated taking account of changes in thermal properties of MOX fuels at higher plutonium contents. The results showed that the heat transfer from the MOX fuel would be slightly smaller than that from UO{sub 2} fuel fragments at similar size in a short period to cause the water hammer. Therefore, the destructive forces were not expected to increase in the new tests with higher plutonium content MOX fuels. (author)

  20. A review of the high temperature creep in oxide nuclear fuels (I)

    International Nuclear Information System (INIS)

    Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C. Y.

    1998-06-01

    Since the initial stage of fuel developmental until recently, considerable efforts have been extensively directed at studying the creep properties of uranium dioxide and its related phases largely due to the importance of their application to the reactor fuels. In this state-of-the-art report, the creep behavior and mechanisms of UO 2 and its related phases were reviewed and discussed in terms of experimental variables such as applied stress, temperature, microstructure and stoichiometry. The objective of this review is to obtain a complete understanding of the influences of these variables on the creep property and creep mechanism in these materials aiming at devising more proper methods for the improvement of the behavior. The database obtained from the results will be primarily utilized also, as the reference data for studying the creep behavior of UO 2 -based mixed oxide nuclear fuels. (author). 64 refs., 6 tabs., 25 figs

  1. Selective oxidation of benzene and cyclohexane using amorphous microporous mixed oxides; Selektive Oxidation von Benzol und Cyclohexan mit amorphen mikroporoesen Mischoxiden

    Energy Technology Data Exchange (ETDEWEB)

    Stoeckmann, M.

    2000-07-01

    Phenol was to be produced by direct oxidation of benzene with environment-friendly oxidants like hydrogen peroxide, oxygen, or ozone. Catalysts were amorphous microporous mixed oxides whose properties can be selected directly in the sol-gel synthesis process. Apart from benzene, also cyclohexane was oxidized with ozone using AMM catalysts in order to get more information on the potential of ozone as oxidant in heterogeneously catalyzed reactions. [German] Ziel dieser Arbeit war die Herstellung von Phenol durch die Direktoxidation von Benzol mit umweltfreundlichen Oxidationsmitteln wie Wasserstoffperoxid, Sauerstoff oder Ozon. Als Katalysatoren dienten amorphe mikroporoese Mischoxide, da deren Eigenschaften direkt in der Synthese durch den Sol-Gel-Prozess gezielt eingestellt werden koennen. Neben Benzol wurde auch Cyclohexan mit Ozon unter der Verwendung von AMM-Katalysatoren oxidiert, um das Potential von Ozon als Oxiationsmittel in heterogen katalysierten Reaktionen naeher zu untersuchen. (orig.)

  2. Physical Properties of Biomass Fuel Briquette from Oil Palm Residues

    African Journals Online (AJOL)

    Palm Kernel Shell (PKS) and Mesocarp Fibre (MF) were used for the production of fuel briquettes in this study in order to supplement the energy mix of the nation. PKS was pulverized and then sieved into different grain particles of 350 μm, 250 μm and 150 μm, before mixing with MF in the ratios: 90:10, 80:20 and 70:30 ...

  3. Biocidal properties of anti-icing additives for aircraft fuels.

    Science.gov (United States)

    Neihof, R A; Bailey, C A

    1978-04-01

    The biocidal and biostatic activities of seven glycol monoalkyl ether compounds were evaluated as part of an effort to find an improved anti-icing additive for jet aircraft fuel. Typical fuel contaminants, Cladosporium resinae, Gliomastix sp., Candida sp., Pseudomonas aeruginosa, and a mixed culture containing sulfate-reducing bacteria were used as assay organisms. Studies were carried out over 3 to 4 months in two-phase systems containing jet fuel and aqueous media. Diethylene glycol monomethyl ether, triethylene glycol monomethyl ether, triethylene glycol monoethyl ether, and 2-methoxyethanol were generally biocidal in aqueous concentrations of 10 to 17% for all organisms except Gliomastix, which required 25% or more. 2-Ethoxyethanol, 2-propoxyethanol, and 2-butoxyethanol were biocidal at progressively lower concentrations down to 1 to 2% for 2-butoxyethanol. The enhanced antimicrobial activity of these three compounds was attributed to cytoplasmic membrane damage because of the correlation between surface tension measurements and lytic activity with P. aeruginosa cells. The mechanism of action of the less active compounds appeared to be due to osmotic (dehydrating) effects. When all requirements are taken into account, diethylene glycol monomethyl ether appears to be the most promising replacement for the currently used additive, 2-methoxyethanol.

  4. Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

    International Nuclear Information System (INIS)

    Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

    2009-01-01

    To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

  5. Production, characterization and fuel properties of alternative diesel fuel from pyrolysis of waste plastic grocery bags

    Science.gov (United States)

    Pyrolysis of HDPE waste grocery bags followed by distillation resulted in a liquid hydrocarbon mixture that consisted of saturated aliphatic paraffins (96.8%), aliphatic olefins (2.6%), and aromatics (0.6%) that corresponded to the boiling range of conventional petroleum diesel fuel (#1 diesel 182–2...

  6. Molecular-Dynamic Simulation In Substation Of Advanced Fuel With Improved Properties

    Energy Technology Data Exchange (ETDEWEB)

    Kolokol, Alexander S.; Shimkevich, Alexander L. [Russian Research Center ' Kurchatov Institute' , 1 Kurchatov Sq. Moscow 123182 (Russian Federation)

    2008-07-01

    A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO{sub 2}+Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic structure improving its thermal and physical properties is formulated here. (authors)

  7. Molecular-Dynamic Simulation In Substation Of Advanced Fuel With Improved Properties

    International Nuclear Information System (INIS)

    Kolokol, Alexander S.; Shimkevich, Alexander L.

    2008-01-01

    A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO 2 +Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic structure improving its thermal and physical properties is formulated here. (authors)

  8. Thoria-based nuclear fuels thermophysical and thermodynamic properties, fabrication, reprocessing, and waste management

    CERN Document Server

    Bharadwaj, S R

    2013-01-01

    This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...

  9. Properties of U sub 3 O sub 8 -aluminum cermet fuel

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B.

    1989-10-01

    Nuclear fuel elements containing U{sub 3}O{sub 8} dispersed in an aluminum matrix have been used in research and test reactors for about 30 years. These elements, sometimes called cermet fuel, are made by powder metallurgical methods (PM) and can accommodate up to approximately 50 wt % uranium in the core section of extruded tubes. Cermet fuel elements have been fabricated and irradiated at the Savannah River Site (SRS). Irradiation behavior is excellent. Extruded tubes with up to 50 wt % uranium have been successfully irradiated to fission densities of about 2 {times} 10{sup 21} fissions per cc of core. Physical, mechanical, and chemical properties of cermet fuels are assembled into a reference document. Results will be used by Argonne National Laboratory to design cermet fuel elements for possible use in the New Production Reactor at SRS. 57 refs., 33 figs., 12 tabs.

  10. Synergy of FexCe1−xO2 mixed oxides for N2O decomposition

    NARCIS (Netherlands)

    Perez-Alonso, F.J.; Melián-Cabrera, I.; López Granados, M.; Kapteijn, F.; Fierro, J.L.G.

    2006-01-01

    Fe–Ce mixed oxides prepared by coprecipitation showed considerable synergy in N2O decomposition when compared with pure metal oxide counterparts. The mixed system also displayed higher stability in reaction at high temperature. Through characterisation by XRD, XPS and TPR, the activity–stability

  11. Synergy of FexCe1-xO2 mixed oxides for N2O decomposition

    NARCIS (Netherlands)

    Perez-Alonso, FJ; Melian Cabrera, Ignacio; Granados, ML; Kapteijn, F; Fierro, JLG

    2006-01-01

    Fe-Ce mixed oxides prepared by coprecipitation showed considerable synergy in N2O decomposition when compared with pure metal oxide counterparts. The mixed system also displayed higher stability in reaction at high temperature. Through characterisation by XRD, XPS and TPR, the activity-stability

  12. Ni–Ta–O mixed oxide catalysts for the low temperature oxidative dehydrogenation of ethane to ethylene

    KAUST Repository

    Zhu, Haibo; Rosenfeld, Devon C.; Anjum, Dalaver H.; Sangaru, Shiv; Saih, Youssef; Ould-Chikh, Samy; Basset, Jean-Marie

    2015-01-01

    The "wet" sol-gel and "dry" solid-state methods were used to prepare Ni-Ta-O mixed oxide catalysts. The resulting Ni-Ta oxides exhibit high activity and selectivity for the low temperature oxidative dehydrogenation of ethane to ethylene. The Ta

  13. Optimization of CeO2-ZrO2 mixed oxide catalysts for ethyl acetate combustion

    Czech Academy of Sciences Publication Activity Database

    Dimitrov, M.; Ivanova, R.; Štengl, Václav; Henych, Jiří; Kovacheva, D.; Tsoncheva, T.

    2015-01-01

    Roč. 47, č. 1 (2015), s. 323-329 ISSN 0324-1130 Institutional support: RVO:61388980 Keywords : nanosized CeO2-ZrO2 * mixed oxide phase * ethyl acetate combustion Subject RIV: CA - Inorganic Chemistry Impact factor: 0.229, year: 2015

  14. Synthesis of glycerol carbonate by transesterification of glycerol with dimethyl carbonate over MgAl mixed oxide catalysts

    NARCIS (Netherlands)

    Liu, P.; Derchi, M.; Hensen, E.J.M.

    2013-01-01

    A series of hydrotalcite-like layered double hydroxides (LDHx) with different Mg/Al atomic ratios (x = 2–6) were prepared by using the co-precipitation method. Further calcination yields mixed oxides with tunable basicity. The basicity of the calcined LDHx (LDOx) strongly depends on the Mg/Al ratio

  15. XPS Spectra Analysis of Ti2+, Ti3+ Ions and Dye Photodegradation Evaluation of Titania-Silica Mixed Oxide Nanoparticles

    Science.gov (United States)

    Chinh, Vu Duc; Broggi, Alessandra; Di Palma, Luca; Scarsella, Marco; Speranza, Giorgio; Vilardi, Giorgio; Thang, Pham Nam

    2018-04-01

    TiO2-SiO2 mixed oxides have been prepared by the sol-gel technique from tetrabutyl orthotitanate and tetraethyl orthosilicate. The prepared materials were characterized by x-ray diffraction, scanning electron microscopy, energy dispersive x-ray spectroscopy, nitrogen physisorption, Fourier-transform infrared spectroscopy (FT-IR) and x-ray photoelectron spectroscopy (XPS). The results indicate that the TiO2-SiO2 mixed oxides have a large surface area and a nanoscale size. FT-IR spectra show that Ti atoms are bonded to silica by oxygen bridging atoms in Ti-O-Si bonds. The titanium valence states in TiO2-SiO2 mixed oxides were investigated by XPS, and their spectra report the presence of Ti2+ and Ti3+ cations for high silica concentration, suggesting the formation of oxygen vacancies. The photocatalytic activity of the prepared materials has been evaluated for the photodegradation of methylene blue (MB). The mixed oxides were activated by means of a UV light source, and the concentration of MB was monitored by UV-Vis spectroscopy. The synthesized TiO2-SiO2 shows significantly higher MB removal efficiency in comparison with that of the commercial TiO2 Degussa, P25.

  16. Preparation of catalysts based on Ce-Mn mixed oxide by coprecipitation for combustion of n-hexane

    International Nuclear Information System (INIS)

    Picasso, Gino; Zavala, Cesar; Cruz, Romulo; Sun Kou, Rosario; Lopez, Alcides

    2013-01-01

    Catalysts based on Ce-Mn mixed with different Ce/Mn molar ratios ranging from 0,5 to 2 have been prepared by coprecipitation at pH constant with ageing times of 4, 18 and 24 h for combustion of n-hexane. XRD patterns of the mixed oxides showed the majority presence of fluorite phase. Specific BET surface areas of mixed oxides were always higher than their single counterparts and their adsorption isotherm depicted a mesoporous surface of Type IV. TPR thermograms confirmed the presence of mixed oxide phase, whose profile shifted to smaller temperatures with increasing content of ceria. Catalytic tests were performed with 2000 ppm of n-hexane and WHSV of 80 h -1 in a fixed-bed reactor. For all samples, only CO 2 and water were observed at total conversion and no partial combustion products were obtained. Ce-Mn mixed oxides were more active than simple oxide samples no matter the aging time. Mixed samples presented thermal stability in contrast with simple ones. Mixed sample with Ce/Mn molar ratio of 2 depicted the highest activity probably due to higher surface area and better reducibility ability of mixed phase. (author)

  17. Impact of thermodynamic properties and heat loss on ignition of transportation fuels in rapid compression machines

    KAUST Repository

    Ahmed, Ahfaz

    2018-01-30

    Rapid compression machines (RCM) are extensively used to study autoignition of a wide variety of fuels at engine relevant conditions. Fuels ranging from pure species to full boiling range gasoline and diesel can be studied in an RCM to develop a better understanding of autoignition kinetics in low to intermediate temperature ranges. In an RCM, autoignition is achieved by compressing a fuel/oxidizer mixture to higher pressure and temperature, thereby initiating chemical reactions promoting ignition. During these experiments, the pressure is continuously monitored and is used to deduce significant events such as the end of compression and the onset of ignition. The pressure profile is also used to assess the temperature evolution of the gas mixture with time using the adiabatic core hypothesis and the heat capacity ratio of the gas mixture. In such RCM studies, real transportation fuels containing many components are often represented by simpler surrogate fuels. While simpler surrogates such as primary reference fuels (PRFs) and ternary primary reference fuel (TPRFs) can match research and motor octane number of transportation fuels, they may not accurately replicate thermodynamic properties (including heat capacity ratio). This non-conformity could exhibit significant discrepancies in the end of compression temperature, thereby affecting ignition delay (τign) measurements. Another aspect of RCMs that can affect τign measurement is post compression heat loss, which depends on various RCM parameters including geometry, extent of insulation, pre-heating temperature etc. To, better understand the effects of these non-chemical kinetic parameters on τign, thermodynamic properties of a number of FACE G gasoline surrogates were calculated and simulated in a multi-zone RCM model. The problem was further investigated using a variance based analysis and individual sensitivities were calculated. This study highlights the effects on τign due to thermodynamic properties of

  18. Impact of thermodynamic properties and heat loss on ignition of transportation fuels in rapid compression machines

    KAUST Repository

    Ahmed, Ahfaz; Hantouche, Mireille; Khurshid, Muneeb; Mohamed, Samah; Nasir, Ehson Fawad; Farooq, Aamir; Roberts, William L.; Knio, Omar; Sarathy, Mani

    2018-01-01

    Rapid compression machines (RCM) are extensively used to study autoignition of a wide variety of fuels at engine relevant conditions. Fuels ranging from pure species to full boiling range gasoline and diesel can be studied in an RCM to develop a better understanding of autoignition kinetics in low to intermediate temperature ranges. In an RCM, autoignition is achieved by compressing a fuel/oxidizer mixture to higher pressure and temperature, thereby initiating chemical reactions promoting ignition. During these experiments, the pressure is continuously monitored and is used to deduce significant events such as the end of compression and the onset of ignition. The pressure profile is also used to assess the temperature evolution of the gas mixture with time using the adiabatic core hypothesis and the heat capacity ratio of the gas mixture. In such RCM studies, real transportation fuels containing many components are often represented by simpler surrogate fuels. While simpler surrogates such as primary reference fuels (PRFs) and ternary primary reference fuel (TPRFs) can match research and motor octane number of transportation fuels, they may not accurately replicate thermodynamic properties (including heat capacity ratio). This non-conformity could exhibit significant discrepancies in the end of compression temperature, thereby affecting ignition delay (τign) measurements. Another aspect of RCMs that can affect τign measurement is post compression heat loss, which depends on various RCM parameters including geometry, extent of insulation, pre-heating temperature etc. To, better understand the effects of these non-chemical kinetic parameters on τign, thermodynamic properties of a number of FACE G gasoline surrogates were calculated and simulated in a multi-zone RCM model. The problem was further investigated using a variance based analysis and individual sensitivities were calculated. This study highlights the effects on τign due to thermodynamic properties of

  19. Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho; Lee, Young-woo; Cho, Moon-sung

    2015-01-01

    High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

  20. Numerical Investigation on Sensitivity of Liquid Jet Breakup to Physical Fuel Properties with Experimental Comparison

    Science.gov (United States)

    Kim, Dokyun; Bravo, Luis; Matusik, Katarzyna; Duke, Daniel; Kastengren, Alan; Swantek, Andy; Powell, Christopher; Ham, Frank

    2016-11-01

    One of the major concerns in modern direct injection engines is the sensitivity of engine performance to fuel characteristics. Recent works have shown that even slight differences in fuel properties can cause significant changes in efficiency and emission of an engine. Since the combustion process is very sensitive to the fuel/air mixture formation resulting from disintegration of liquid jet, the precise assessment of fuel sensitivity on liquid jet atomization process is required first to study the impact of different fuels on the combustion. In the present study, the breaking process of a liquid jet from a diesel injector injecting into a quiescent gas chamber is investigated numerically and experimentally for different liquid fuels (n-dodecane, iso-octane, CAT A2 and C3). The unsplit geometric Volume-of-Fluid method is employed to capture the phase interface in Large-eddy simulations and results are compared against the radiography measurement from Argonne National Lab including jet penetration, liquid mass distribution and volume fraction. The breakup characteristics will be shown for different fuels as well as droplet PDF statistics to demonstrate the influences of the physical properties on the primary atomization of liquid jet. Supported by HPCMP FRONTIER award, US DOD, Office of the Army.

  1. Alcohols/Ethers as Oxygenates in Diesel Fuel: Properties of Blended Fuels and Evaluation of Practiacl Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Nylund, N.; Aakko, P. [TEC Trans Energy Consulting Ltd (Finland); Niemi, S.; Paanu, T. [Turku Polytechnic (Finland); Berg, R. [Befri Konsult (Sweden)

    2005-03-15

    Oxygenates blended into diesel fuel can serve at least two purposes. Components based on renewable feedstocks make it possible to introduce a renewable component into diesel fuel. Secondly, oxygenates blended into diesel fuel might help to reduce emissions. A number of different oxygenates have been considered as components for diesel fuel. These oxygenates include various alcohols, ethers, esters and carbonates. Of the oxygenates, ethanol is the most common and almost all practical experiences have been generated from the use of diesel/ethanol blends (E-diesel). Biodiesel was not included in this study. Adding ethanol to diesel will reduce cetane, and therefore, both cetane improver and lubricity additives might be needed. Diesel/ethanol emulsions obtained with emulsifiers or without additives are 'milky' mixtures. Micro-emulsions of ethanol and diesel can be obtained using additives containing surfactants or co-solvents. The microemulsions are chemically and thermodynamically stable, they are clear and bright blends, unlike the emulsions. Storage and handling regulations for fuels are based on the flash point. The problem with, e.g., ethanol into diesel is that ethanol lowers the flash point of the blend significantly even at low concentrations. Regarding safety, diesel-ethanol blends fall into the same category as gasoline. Higher alcohols are more suitable for diesel blending than ethanol. Currently, various standards and specifications set rather tight limits for diesel fuel composition and properties. It should be noted that, e.g., E-diesel does not fulfil any current diesel specification and it cannot, thus, be sold as general diesel fuel. Some blends have already received approvals for special applications. The critical factors of the potential commercial use of these blends include blend properties such as stability, viscosity and lubricity, safety and materials compatibility. The effect of the fuel on engine performance, durability and emissions

  2. Alcohols/Ethers as Oxygenates in Diesel Fuel: Properties of Blended Fuels and Evaluation of Practiacl Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Nylund, N; Aakko, P [TEC Trans Energy Consulting Ltd (Finland); Niemi, S; Paanu, T [Turku Polytechnic (Finland); Berg, R [Befri Konsult (Sweden)

    2005-03-15

    Oxygenates blended into diesel fuel can serve at least two purposes. Components based on renewable feedstocks make it possible to introduce a renewable component into diesel fuel. Secondly, oxygenates blended into diesel fuel might help to reduce emissions. A number of different oxygenates have been considered as components for diesel fuel. These oxygenates include various alcohols, ethers, esters and carbonates. Of the oxygenates, ethanol is the most common and almost all practical experiences have been generated from the use of diesel/ethanol blends (E-diesel). Biodiesel was not included in this study. Adding ethanol to diesel will reduce cetane, and therefore, both cetane improver and lubricity additives might be needed. Diesel/ethanol emulsions obtained with emulsifiers or without additives are 'milky' mixtures. Micro-emulsions of ethanol and diesel can be obtained using additives containing surfactants or co-solvents. The microemulsions are chemically and thermodynamically stable, they are clear and bright blends, unlike the emulsions. Storage and handling regulations for fuels are based on the flash point. The problem with, e.g., ethanol into diesel is that ethanol lowers the flash point of the blend significantly even at low concentrations. Regarding safety, diesel-ethanol blends fall into the same category as gasoline. Higher alcohols are more suitable for diesel blending than ethanol. Currently, various standards and specifications set rather tight limits for diesel fuel composition and properties. It should be noted that, e.g., E-diesel does not fulfil any current diesel specification and it cannot, thus, be sold as general diesel fuel. Some blends have already received approvals for special applications. The critical factors of the potential commercial use of these blends include blend properties such as stability, viscosity and lubricity, safety and materials compatibility. The effect of the fuel on engine performance, durability and emissions is also

  3. Effects of Fuel Oil on the Geotechnical Properties of Clay Soil

    Directory of Open Access Journals (Sweden)

    Mahdi Obaid Karkush

    2017-08-01

    Full Text Available The present study highlights the effects of medium fuel oil (MFO on the chemical, physical and mechanical properties of clay soil samples (disturbed and undisturbed obtained from the site of the electrical power plant in the campus of the University of Baghdad at Al-Jadriah district in Baghdad/Iraq. The soil sample was classified according to the unified soil classification system (USCS as CL and described as lean clay of low plasticity. The medium fuel oil is an industrial wastewater disposed as a byproduct from the fuel used in the electricity power plant. The soil samples are artificially contaminated with two percentages of medium fuel oil, 10 and 20 % related to the dry weight of soil. The soil samples were mixed with the contaminant (MFO by hand and then left for 4 days for homogeneity. A series of laboratory tests are conducted on both natural and artificially contaminated soil samples to measure the effects of medium fuel oil on the chemical, physical and mechanical properties of soil samples. The results of tests showed that the medium fuel oil has significant impacts on some properties of soil and slight effects on the others. Increasing the percentage of contaminant causes a slight decrease in the liquid limit and particle size distribution; on the other hand, it causes a considerable increase in the consolidation parameters and decrease in shear strength parameters. Also, there is a slight change in the chemical composition of soil samples.

  4. Characterization of copper-zinc mixed oxide system in relation to different precursor structure and morphology

    Energy Technology Data Exchange (ETDEWEB)

    Porta, P; De Rossi, S; Ferraris, G [Centro del CNR su ' Struttura e Attivia Catalitica di Sistemi di Ossidi' (SACSO), Rome (Italy); Pompa, F [ENEA, TIB Scienza dei Materiali, Rome (Italy)

    1991-03-01

    Hydroxycarbonate and hydroxynitrate precursors of CuO-ZnO catalysts (Cu/Zn atomic ratio=67/33) have been prepared by two different methods; the precursor obtained by precipitation at 333 K and constant pH=8 from mixed nitrate solution with excess of sodium bicarbonate consisted of zincian malachite and aurichalcite, while that obtained by addition of sodium carbonate solution to Cu-Zn nitrate solution is essentially copper hydroxynitrate plus some amount of aurichalcite. By thermal decomposition at 623 K both types of precursor gave a mixture of CuO and ZnO. The mixed oxides were then treated at 873, 1073 and 1273 K in air. X-ray diffraction, diffuse reflectance spectroscopy, scanning electron microscopy and surface area determination were used to characterize the mixed oxide systems. The precursor containing zincian malachite plus aurichalcite, after calcination at 623 K gave rise to well dispersed and much smaller particles of CuO and ZnO than the precursor containing copper hydroxynitrate plus aurichalcite. No Cu{sup 2+} in solid solution in the ZnO zincite structure Zn{sup 2+} in the CuO tenorite lattice were detected by reflectance spectroscopy up to 873 K; the presence of tetrahedral copper (Cu{sub x}Zn{sub 1-x}O solid solution formation at least at the surface) was evident only in samples calcined at temperatures higher than 1073 K. X-ray diffraction analysis for lattice parameter determination showed that only for samples treated at 1273 K both Cu{sub x}Zn{sub 1-x}O and Zn{sub y}Cu{sub 1-y}O solid solution formations are detectable. An unexpected volume decrease of Zn{sub y}Cu{sub 1-y}O with respect to pure CuO was revealed; the introduction of zinc in the tenorite structure probably changes the local metal symmetry from nearly square planar towards octahedral, producing an overall less distorted and more compact structure. (orig.).

  5. Structural and microstructural changes in the zirconium-indium mixed oxide system during the thermal treatment

    Science.gov (United States)

    Štefanić, G.; Štefanić, I. I.; Musić, S.; Ivanda, M.

    2011-05-01

    The zirconium-indium mixed oxide systems on both the zirconium- and the indium-rich side of the concentration range were prepared by co-precipitation from aqueous solutions of the corresponding salts, followed by washing and heat-treatment. The thermal behavior (up to 1000 °C) of the dried samples was examined by X-ray powder diffraction, Raman spectroscopy, Fourier transform infrared spectroscopy, field emission scanning electron microscopy, energy dispersive X-ray spectrometry, differential thermal analysis and thermogravimetric measurements. The obtained results show that the increase in the amount of the second phase causes an increase of both the crystallization temperature of the amorphous precursors of ZrO 2, from 435 °C (0 mol.% of InO 1.5) to 476 °C (˜62 mol.% of InO 1.5), and of the topotactic transition temperature of cubic In(OH) 3 to cubic In 2O 3, from 259 °C (0 mol.% of ZrO 2) to 290 °C (˜25 mol.% of ZrO 2). The amorphous precursors of ZrO 2 phase exhibit an extended capability to incorporate In 3+ ions (more than 60 mol.%). With a rise in temperature the maximum solubility of In 3+ ions in the ZrO 2 lattice decreases from ˜55 mol.% in the crystallization products obtained after calcination at 400 °C to ˜10 mol.% after calcination at 1000 °C. The results of phase analysis indicate that the incorporation of In 3+ ions partially stabilized both the tetragonal and cubic ZrO 2 polymorphs. The maximum solubility of Zr 4+ ions in the starting In(OH) 3 lattice was estimated at ˜10 mol.%. Thermal treatment causes a small increase of Zr 4+ ion solubility limits, estimated at ˜15 mol.% in the cubic In 2O 3 lattice after calcination at 1000 °C. Precise lattice parameter measurements, by using Le Bail refinements of the powder diffraction patterns with added silicon as an internal standard, show that the incorporation of In 3+ ions caused a very small decrease of the cubic ZrO 2 lattice, while the incorporation of Zr 4+ ions had a negligible

  6. Selective oxidation of n-butane to maleic anhydride under oxygen-deficient conditions over V-P-O mixed oxides

    NARCIS (Netherlands)

    Bosch, H.; Bruggink, A.A.; Ross, J.R.H.

    1987-01-01

    The selective oxidation of n-butane to maleic anhydride over V-P-O mixed oxides was studied under oxygen deficient conditions. The mixed oxides were prepared with P/V atomic ratios ranging from 0.7 to 1.0. Catalysts with P/V <1.0 did not show any selectivity to maleic anhydride formation, regardless

  7. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    1999-01-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  8. Mixed oxidizer hybrid propulsion system optimization under uncertainty using applied response surface methodology and Monte Carlo simulation

    Science.gov (United States)

    Whitehead, James Joshua

    The analysis documented herein provides an integrated approach for the conduct of optimization under uncertainty (OUU) using Monte Carlo Simulation (MCS) techniques coupled with response surface-based methods for characterization of mixture-dependent variables. This novel methodology provides an innovative means of conducting optimization studies under uncertainty in propulsion system design. Analytic inputs are based upon empirical regression rate information obtained from design of experiments (DOE) mixture studies utilizing a mixed oxidizer hybrid rocket concept. Hybrid fuel regression rate was selected as the target response variable for optimization under uncertainty, with maximization of regression rate chosen as the driving objective. Characteristic operational conditions and propellant mixture compositions from experimental efforts conducted during previous foundational work were combined with elemental uncertainty estimates as input variables. Response surfaces for mixture-dependent variables and their associated uncertainty levels were developed using quadratic response equations incorporating single and two-factor interactions. These analysis inputs, response surface equations and associated uncertainty contributions were applied to a probabilistic MCS to develop dispersed regression rates as a function of operational and mixture input conditions within design space. Illustrative case scenarios were developed and assessed using this analytic approach including fully and partially constrained operational condition sets over all of design mixture space. In addition, optimization sets were performed across an operationally representative region in operational space and across all investigated mixture combinations. These scenarios were selected as representative examples relevant to propulsion system optimization, particularly for hybrid and solid rocket platforms. Ternary diagrams, including contour and surface plots, were developed and utilized to aid in

  9. The manufacture process and properties of (U, Gd)O2 burnable poisonous fuel pellets

    International Nuclear Information System (INIS)

    Yi Wei; Tang Yueming; Dai Shengping; Yang Youqing; Zuo Guoping; Wu Shihong; Gu Xiaofei; Gu Mingfei

    2006-03-01

    The main properties of important raw powder materials used in the (U, Gd)O 2 burnable poisonous fuel pellets production line of NPIC are presented. The powders included UO 2 , Gd 2 O 3 , (U, Gd) 3 O 8 and necessary additives, such as ammonium oxalate and zinc stearate. And the main properties of (U, Gd)O 2 burnable poisonous fuel pellets and the manufacture processes, such as ball-milling blending, granulation, pressing, sintering and grinding are also described. Moreover, the main effect of the process parameters controlled in the manufacture process have been discussed. (authors)

  10. Compound process fuel cycle concept

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo

    2005-01-01

    Mass flow of light water reactor spent fuel for a newly proposed nuclear fuel cycle concept 'Compound Process Fuel Cycle' has been studied in order to assess the capacity of the concept for accepting light water reactor spent fuels, taking an example for boiling water reactor mixed oxide spent fuel of 60 GWd/t burn-up and for a fast reactor core of 3 GW thermal output. The acceptable heavy metal of boiling water reactor mixed oxide spent fuel is about 3.7 t/y/reactor while the burn-up of the recycled fuel is about 160 GWd/t and about 1.6 t/y reactor with the recycled fuel burn-up of about 300 GWd/t, in the case of 2 times recycle and 4 times recycle respectively. The compound process fuel cycle concept has such flexibility that it can accept so much light water reactor spent fuels as to suppress the light water reactor spent fuel pile-up if not so high fuel burn-up is expected, and can aim at high fuel burn-up if the light water reactor spent fuel pile-up is not so much. Following distinctive features of the concept have also been revealed. A sort of ideal utilization of boiling water reactor mixed oxide spent fuel might be achieved through this concept, since both plutonium and minor actinide reach equilibrium state beyond 2 times recycle. Changes of the reactivity coefficients during recycles are mild, giving roughly same level of reactivity coefficients as the conventional large scale fast breeder core. Both the radio-activity and the heat generation after 4 year cooling and after 4 times recycle are less than 2.5 times of those of the pre recycle fuel. (author)

  11. Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

    International Nuclear Information System (INIS)

    Chatterjee, S.

    2005-01-01

    During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

  12. Catalytic combustion of methane over mixed oxides derived from Co-Mg/Al ternary hydrotalcites

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Zheng [Inorganic Chemistry Laboratory, Department of Chemistry, University of Oxford, OX1 3QR (United Kingdom); Research Centre of Eco-Environmental Sciences, CAS, Beijing 100085 (China); Jesus College, University of Oxford, OX1 3DW (United Kingdom); Yu, Junjie; Cheng, Jie; Hao, Zhengping [Research Centre of Eco-Environmental Sciences, CAS, Beijing 100085 (China); Xiao, Tiancun; Edwards, Peter P. [Inorganic Chemistry Laboratory, Department of Chemistry, University of Oxford, OX1 3QR (United Kingdom); Jones, Martin O. [Inorganic Chemistry Laboratory, Department of Chemistry, University of Oxford, OX1 3QR (United Kingdom); Rutherford Appleton Laboratory, Didcot, OX11 0QX (United Kingdom)

    2010-01-15

    Co{sub x}Mg{sub 3-x} /Al composite oxides (xCoMAO-800) were prepared by calcination of Co{sub x}Mg{sub 3-x}/Al hydrotalcites (x=0.0,0.5,1.0,1.5,2.0,2.5,3.0, respectively) at 800 C. The materials were characterized using XRD, TG-DSC, N{sub 2} adsorption-desorption and TPR. The methane catalytic combustion over the xCoMAO-800 was assessed in a fixed bed micro-reactor. The results revealed that cobalt can be homogenously dispersed into the matrices of the hydrotalcites and determines the structure, specific surface areas and porosity of the derived xCoMAO-800 oxide catalysts. The thermal stability and homogeneity of the hydrotalcites markedly depends on the cobalt concentration in the hydrotalcites. The Co-based hydrotalcite-derived oxides exhibit good activity in the catalytic combustion of methane. The catalytic activity over the xCoMAO-800 oxides enhances with increasing x up to 1.5, but subsequently decreases dramatically as cobalt loadings are further increased. The 1.5CoMAO-800 catalyst shows the best methane combustion activity, igniting methane at 450 C and completing methane combustion around 600 C. The catalytic combustion activity over the xCoMAO-800 oxides are closely related to the strong Co-Mg/Al interaction within the mixed oxides according to the TG-DSC, TPR and activity characteristics. (author)

  13. Determining the optimum conditions for modified diesel fuel combustion considering its emission, properties and engine performance

    International Nuclear Information System (INIS)

    Fayyazbakhsh, Ahmad; Pirouzfar, Vahid

    2016-01-01

    Highlights: • Gas emissions, fuel properties and performance engine modeling. • Optimization of new modified fuel prepared from n-Butanol and Nano particles. • Model accuracy analysis. - Abstract: This essay scrutinizes an experimental study conducted to appraise the influence of using n-Butanol with diesel fuel in 5% and 10% (volume) n-Butanol, 1% nitro methane (NM), injection timing and two Nano-particles (alumina and a type of silica powder) on the engine performance (brake specific fuel consumption and engine power), fuel properties (Cetane number and flash point) and exhaust emissions (soot, NO_x and CO) of an engine with 4-cylinder (with a system of common rail fuel injection), intercooling, cooled exhaust gas recirculation (EGR), and turbocharged. The tests are conducted by varying the engine load (25 and 75 nm) and changing engine speed (1500 and 2200 rpm). Normal Butanol presents better brake specific fuel consumption (BSFC) but this blend doesn’t reflect better engine power. All the percentages of n-Butanol in the fuel make Cetane number decrease but adding 1% of nitro methane makes Cetane number increase. For all the n-Butanol, the percentage flash makes the fuel decrease in comparison to pure diesel fuel. The current experimental study demonstrates that adding the n-Butanol and nitro methane to diesel fuel direct into diminishing soot emission. In contrast, this blend raises NO_x and CO emissions. Furthermore, this research indicates that the increase of engine speed dwindle air pollutants and enhances BSFC. It also remarks that power gets increased at low engine speed. However, power gets reducedat high speed. This article represents that the increasing of engine load leads to increasing all of air pollutant, increasing of power and decreasing of brake specific fuel consumption. Both the Cetane number and flash point are independent from engine speed and engine load. The present paper shows that the effect of silica with high percentage of n

  14. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

  15. The low enriched uranium fuel cycle in Ontario

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-02-01

    Six fuel-cycle strategies for use in CANDU reactors are examined in terms of their uranium-conserving properties and their ease of commercialization for three assumed growth rates of installed nuclear capacity in Ontario. The fuel cycle strategies considered assume the continued use of the natural uranium cycle up to the mid-1990's. At that time, the low-enriched uranium (LEU) cycle is gradually introduced into the existing power generation grid. In the mid-2020's one of four advanced cycles is introduced. The advanced cycles considered are: mixed oxide, intermediate burn-up thorium (Pu topping), intermediate burn-up thorium (U topping), and LMFBR. For comparison purposes an all natural uranium strategy and a natural uranium-LEU strategy (with no advanced cycle) are also included. None of the strategies emerges as a clear, overall best choice. (LL)

  16. Synthesis of ZnO nanowire arrays on ZnO−TiO{sub 2} mixed oxide seed layer for dye sensitized solar cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Marimuthu, T. [Advanced Materials and Thin Film Physics Lab, Department of Physics, Alagappa University, Karaikudi (India); Anandhan, N., E-mail: anandhan_kn@rediffmail.com [Advanced Materials and Thin Film Physics Lab, Department of Physics, Alagappa University, Karaikudi (India); Thangamuthu, R. [Electrochemical Materials Science Division, CSIR-Central Electrochemical Research Institute, Karaikudi (India); Mummoorthi, M. [Advanced Materials and Thin Film Physics Lab, Department of Physics, Alagappa University, Karaikudi (India); Ravi, G. [Photonic Crystal Lab, Department of Physics, Alagappa University, Karaikudi (India)

    2016-08-25

    ZnO nanowire arrays (NWAs) were synthesized on ZnO−TiO{sub 2} mixed oxide seeded FTO conducting glass plate by two-step sol-gel and hydrothermal method, respectively. X-ray diffraction patterns reveal the presence of mixed and hexagonal phases in seed layer and NWAs, respectively. Scanning electron microscope images showed that the FTO glass plate is uniformly covered with grains and a few nanorods in seed layer and dense NWAs are vertically grown on the seed layer. The hexagonal structure and high crystal quality have been confirmed by micro Raman spectra. Photoluminescence spectra also present that NWAs have high crystal quality and less atomic defects. UV spectra indicate that NWAs are absorbed more dye molecules and it has the band gap equal to bulk material. The efficiency of ZnO−TiO{sub 2} mixed oxide seed layer and ZnO NWAs is found to be 0.56 and 0.84% respectively. Electrochemical impedance spectra reveal that NWAs DSSC has high charge transfer recombination resistance than the seed layer DSSC. - Highlights: • ZnO nanowire arrays were synthesized by two-step sol-gel and hydrothermal method. • The crystal structure and crystalline quality of films are confirmed by Raman spectra. • The emission properties of films are investigated by photoluminescence spectra. • ZnO nanowire arrays (NWAs) have higher charge transfer recombination resistance. • The conversion efficiency of the seed layer and NWAs is to be 0.56 and 0.84%.

  17. Effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu K; Malachenko, L L; Medvedev, A V; Solyany, V I; Sukhanov, G I; Tonkov, V Yu

    1987-05-01

    Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. 4 references, 7 figures, 3 tables.

  18. The effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Malachenko, L.L.; Medvedev, A.V.; Solyany, V.I.; Sukhanov, G.I.; Tonkov, V.Yu.

    1987-01-01

    Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    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)

  20. Properties of plasma flames sustained by microwaves and burning hydrocarbon fuels

    International Nuclear Information System (INIS)

    Hong, Yong Cheol; Uhm, Han Sup

    2006-01-01

    Plasma flames made of atmospheric microwave plasma and a fuel-burning flame were presented and their properties were investigated experimentally. The plasma flame generator consists of a fuel injector and a plasma flame exit connected in series to a microwave plasma torch. The plasma flames are sustained by injecting hydrocarbon fuels into a microwave plasma torch in air discharge. The microwave plasma torch in the plasma flame system can burn a hydrocarbon fuel by high-temperature plasma and high atomic oxygen density, decomposing the hydrogen and carbon containing fuel. We present the visual observations of the sustained plasma flames and measure the gas temperature using a thermocouple device in terms of the gas-fuel mixture and flow rate. The plasma flame volume of the hydrocarbon fuel burners was more than approximately 30-50 times that of the torch plasma. While the temperature of the torch plasma flame was only 868 K at a measurement point, that of the diesel microwave plasma flame with the addition of 0.019 lpm diesel and 30 lpm oxygen increased drastically to about 2280 K. Preliminary experiments for methane plasma flame were also carried out, measuring the temperature profiles of flames along the radial and axial directions. Finally, we investigated the influence of the microwave plasma on combustion flame by observing and comparing OH molecular spectra for the methane plasma flame and methane flame only

  1. Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

    International Nuclear Information System (INIS)

    Terrani, K.A.; Kiggans, J.O.; Silva, C.M.; Shih, C.; Katoh, Y.; Snead, L.L.

    2015-01-01

    The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE–SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors

  2. Reference Material Properties and Standard Problems to Verify the Fuel Performance Models Ver 1.0

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, Jae Yong; Koo, Yang Hyun

    2010-12-01

    All fuel performance models must be validated by in-pile and out-pile tests. However, the model validation requires much efforts and times to confirm its exactness. In many fields, new performance models and codes are confirmed by code-to-code benchmarking process under simplified standard problem analysis. At present, the DUOS, which is the steady state fuel performance analysis code for dual cooled annular fuel, development project is progressing and new FEM module is developed to analyze the fuel performance during transient period. In addition, the verification process is planning to examine the new models and module's rightness by comparing with commercial finite element analysis such as a ADINA, ABAQUS and ANSYS. This reports contains the result of unification of material properties and establishment of standard problem to verify the newly developed models with commercial FEM code

  3. SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

    International Nuclear Information System (INIS)

    Dutt, D.S.; Baker, R.B.

    1975-06-01

    The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

  4. Thermochemical properties of media for pyrometallurgical nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Hosoya, Yuji; Terai, Takayuki

    1998-01-01

    Molten chloride/cadmium system is considered to be applied to a solvent in pyrochemical reprocessing of spent nuclear fuel. In this work, phase diagrams for molten chloride systems were constructed, using NdCl 3 as an imitative substance in place of UCl 3 or PuCl 3 . Hastelloy-X (Ni/Cr21/Fe18/Mo9/W) was examined as a structural material for the corrosion-resistance against molten chloride baths containing NdCl 3 . The process of corrosion was thermochemically discussed and the form of the corrosion was illustrated. Rutherford backscattering spectroscopy was successfully applied to determine the elemental distribution profile of specimens tested on the compatibility with molten chloride mixture at elevated temperature. Ferritic steel was also examined as another candidate material for the compatibility with molten cadmium covered with LiCl-KCl eutectic salt. Variation of near-surface composition was observed by comparing the results of Rutherford backscattering spectroscopy obtained before and after the dipping. (author)

  5. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

  6. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

  7. Selective oxidation of naphthalene in CO/H{sub 2} mixtures over Mo/V/W mixed oxides. A contribution to biomass gasification; Selektivoxidation von Naphthalin in CO/H{sub 2}-Mischungen an Mo/V/W-Mischoxiden. Ein Beitrag zur Biomassevergasung

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Sonja

    2007-11-26

    The development of efficient and sustainable technologies for the production of biofuels of the second generation is one of the fundamental challenges of the beginning 21st century. In the first step of the so called Biomass-to-Liquid processes the biomass is transformed to syngas in a gasifier. Prior to fuel generation via Fischer-Tropsch, methanol or dimethyl ether synthesis a purification of the raw gas must take place. A main impurity of the gas is the tar formed during gasification. Besides desactivation of the catalyst in the following synthesis step, the tar condenses during cooling down of the syngas, leading to the formation of scaling in the plant and to a reduced syngas yield. In order to develop a energy efficient technology this work focuses on the development of a method for tar removal that can be performed at temperatures between 400 and 900 C while providing the required syngas purity for the subsequent fuel synthesis. First the potential of a Mo8V2W1Ox mixed oxide catalyst for selective tar oxidation was investigated using a thermo balance. By means of temperature programmed reductions of the oxidised catalyst with CO2, H2 and tar model compound naphthalene no activity regarding the oxidation of CO, only minor activity regarding the oxidation of H2, but high activity for naphthalene oxidation was determined. Based on these studies temperature programmed reactions in the presence of oxygen were performed employing a catalyst bed in a plug flow rector equipped with an online mass spectrometer. The complete conversion of naphthalene was observed at temperatures above 412 C. In the same temperature range no oxidation of CO and only a marginal oxidation of H2 occurred. Apart from the total oxidation products CO, CO2 and H2O partial oxidised products as maleic anhydride and phthalic anhydride were formed, though. Mechanical mixing of the catalyst with sodium carbonate lead to an optimization of the catalytic properties of the mixed oxide catalyst

  8. Storage tank materials for biodiesel blends; the analysis of fuel property changes

    Directory of Open Access Journals (Sweden)

    Nurul Komariah Leily

    2017-01-01

    Full Text Available Fuel stability is one of major problem in biodiesel application. Some of the physical properties of biodiesel are commonly changed during storage. The change in physico-chemical properties is strongly correlated to the stability of the fuel. This study is objected to observe the potential materials for biodiesel storage. The test was conducted in three kinds of tank materials, such as glass, HDPE, and stainless steel. The fuel properties are monitored in 12 weeks, while the sample was analyzed every week. Biodiesel used is palm oil based. The storage tanks were placed in a confined indoor space with range of temperature 27–34 °C. The relative humidity and sunshine duration on the location was also evaluated. The observed properties of the fuel blends were density, viscosity and water content. During 12 weeks of storage, the average density of B20 was changed very slightly in all tanks, while the viscosity was tend to increase sharply, especially in polimerics tank. Water content of B20 was increased by the increase of storage time especially in HDPE tank. In short period of storage, the biodiesel blends is found more stable in glass tank due to its versatility to prohibit oxidation, degradation, and its chemical resistance.

  9. Effects of fuel properties, temperature, and pressure on fuel reactivity, formation and destruction of nitrogen oxides, and release of alkalis

    International Nuclear Information System (INIS)

    Aho, M.

    1998-01-01

    This study assists in the development of advanced combustion technologies (PFBC, IGCC) with high efficiency of electricity production from solid fuels (η = 47 - 50%) and in minimizing emissions of nitrogen oxides in atmospheric and pressurised FB combustion. In addition to the work done within the LIEKKI 2 programme, research work has been carried out inside the Joule 2 programme of EU. The research work may be divided into three parts: (1) Study of N x O y formation and destruction, (2) Study of fuel reactivity at elevated pressures, and (3) Study on alkali release from different coals. Experimental work was carried out utilizing a novel pressurized entrained flow reactor (PEFR) completed in VTT Energy in the autumn 1992. The device was unique in the world between 1992 and 1995. The effects of fuel properties on the formation of N 2 O and NO at conditions typical to FB combustion were studied for a large number of fuels including different coals, coal-derived char, peat, and bark. This work started before 1993 and was completed in 1995. FTIR technology was utilized for on-line gas analysis of N 2 O, NO, and NO 2 . The ratio fuel-O/fuel-N was found to be the most important fuel factor determining the formation of N 2 O and NO from volatile fuel-N. Only a small part of N 2 O is formed from char-N. The effect of pressure (0.2 - 2.0 MPa) on the formation of N 2 O, NO, and NO 2 , and destruction of NO with ammonia (Thermal DeNO x , experiments at 0.2, 0.5, and 1.5 MPa) and urea (NO x Out, experiments at 0.5 MPa) were studied in cooperation with Aabo Akademi University (AaAU). VTT performed the experimental work and AaAU the kinetic modelling. A part of these results are presented in the report by AaAU. Increase of pressure decreases NO formation and increases NO 2 formation. The behaviour of N 2 O is more complex. Both destruction processes for NO seem to operate well at elevated pressure, although clear effects of pressure on the temperature window of Thermal DeNO x

  10. Modeling transit bus fuel consumption on the basis of cycle properties.

    Science.gov (United States)

    Delgado, Oscar F; Clark, Nigel N; Thompson, Gregory J

    2011-04-01

    A method exists to predict heavy-duty vehicle fuel economy and emissions over an "unseen" cycle or during unseen on-road activity on the basis of fuel consumption and emissions data from measured chassis dynamometer test cycles and properties (statistical parameters) of those cycles. No regression is required for the method, which relies solely on the linear association of vehicle performance with cycle properties. This method has been advanced and examined using previously published heavy-duty truck data gathered using the West Virginia University heavy-duty chassis dynamometer with the trucks exercised over limited test cycles. In this study, data were available from a Washington Metropolitan Area Transit Authority emission testing program conducted in 2006. Chassis dynamometer data from two conventional diesel buses, two compressed natural gas buses, and one hybrid diesel bus were evaluated using an expanded driving cycle set of 16 or 17 different driving cycles. Cycle properties and vehicle fuel consumption measurements from three baseline cycles were selected to generate a linear model and then to predict unseen fuel consumption over the remaining 13 or 14 cycles. Average velocity, average positive acceleration, and number of stops per distance were found to be the desired cycle properties for use in the model. The methodology allowed for the prediction of fuel consumption with an average error of 8.5% from vehicles operating on a diverse set of chassis dynamometer cycles on the basis of relatively few experimental measurements. It was found that the data used for prediction should be acquired from a set that must include an idle cycle along with a relatively slow transient cycle and a relatively high speed cycle. The method was also applied to oxides of nitrogen prediction and was found to have less predictive capability than for fuel consumption with an average error of 20.4%.

  11. Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors; Comparação de ciclos abertos de urânio e óxidos mistos de tório-urânio utilizando reatores avançados

    Energy Technology Data Exchange (ETDEWEB)

    Gonçalves, Letícia C.; Maiorino, José R., E-mail: goncalves.leticiac@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharia, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    A comparative study of the mass balance and production costs of uranium oxide fuels was carried out for an AP1000 reactor and thorium-uranium mixed oxide in a reactor proposal using thorium called AP-Th1000. Assuming the input mass values for a fuel load the average enrichment for both reactors as well as their feed mass was determined. With these parameters, the costs were calculated in each fuel preparation process, assuming the prices provided by the World Nuclear Association. The total fuel costs for the two reactors were quantitatively compared with 18-month open cycle. Considering enrichment of 20% for the open cycle of mixed U-Th oxide fuel, the total uranium consumption of this option was 50% higher and the cost due to the enrichment was 70% higher. The results show that the use of U-Th mixed oxide fuels can be advantageous considering sustainability issues. In this case other parameters and conditions should be investigated, especially those related to fuel recycling, spent fuel storage and reduction of the amount of transuranic radioactive waste.

  12. Removal of Hazardous Pollutants from Wastewaters: Applications of TiO2-SiO2 Mixed Oxide Materials

    Directory of Open Access Journals (Sweden)

    Shivatharsiny Rasalingam

    2014-01-01

    Full Text Available The direct release of untreated wastewaters from various industries and households results in the release of toxic pollutants to the aquatic environment. Advanced oxidation processes (AOP have gained wide attention owing to the prospect of complete mineralization of nonbiodegradable organic substances to environmentally innocuous products by chemical oxidation. In particular, heterogeneous photocatalysis has been demonstrated to have tremendous promise in water purification and treatment of several pollutant materials that include naturally occurring toxins, pesticides, and other deleterious contaminants. In this work, we have reviewed the different removal techniques that have been employed for water purification. In particular, the application of TiO2-SiO2 binary mixed oxide materials for wastewater treatment is explained herein, and it is evident from the literature survey that these mixed oxide materials have enhanced abilities to remove a wide variety of pollutants.

  13. Preparation and characterization of vanadia-titania mixed oxide for immobilization of Serratia rubidaea CCT 5732 and Klebsiella marcescens bacteria

    International Nuclear Information System (INIS)

    Saragiotto Colpini, Leda Maria; Correia Goncalves, Regina A.; Goncalves, Jose Eduardo; Maieru Macedo Costa, Creusa

    2008-01-01

    Vanadia-titania mixed oxide was synthesized by sol-gel method and characterized by several techniques. Texturally, it is formed by mesopores and presents high-specific surface area and controlled porosity. Scanning electron microscopy revealed that vanadium is homogeneously distributed in the material. Structurally, it was possible to identify characteristic V=O stretching bands by IR. The analysis of X-ray diffraction showed that the material, particularly vanadium, is highly dispersed. Application experiments were carried out through the immobilization of Serratia rubidae CCT 5732 and Klebsiella marcescens bacteria by adsorption on the surface of mixed oxide. The micrographies revealed that the bacteria were adsorbed on the entire support, with average surface densities of 8.55 x 10 11 cells/m 2 (Serratia rubidae CCT 5732) and 3.40 x 10 11 cells/m 2 (K. marcescens)

  14. Determination of nitrogen in PuO2-UO2 mixed oxide with fusion-TCD method

    International Nuclear Information System (INIS)

    Hiyama, Toshiaki; Takahashi, Toshio; Ohuchi, Yoshiro; Suzuki, Takeshi; Kaya, Akira

    1983-01-01

    Nitrogen in plutonium and uranium mixed oxide has been determined by several different methods. In the classical Dumas method, nitrogen is measured volumetrically. Other procedures are modifications of the Kjeldahl method in which nitrogen is converted to ammonia. All these methods are accompanied with the same problems in which the analysis takes usually 2 or 3 days with waste solutions generated. So far, the authors have tried to determine nitrogen in PuO 2 -UO 2 mixed oxide with an improved dry process, in which an impulse furnace is used to heat the sample in a graphite crucible under a helium atmosphere. Nitrogen released from the sample were separated from other gases (i.e., a large amount of carbon monoxide, very small amount of hydrogen etc.) by a combination of original two moleculer sieve columns and pre-cut system. Separated nitrogen was then detected on the basis of thermal conductivities. (author)

  15. Solvent-Free Selective Oxidation of Toluene with O2 Catalyzed by Metal Cation Modified LDHs and Mixed Oxides

    Directory of Open Access Journals (Sweden)

    Xiaoli Wang

    2016-01-01

    Full Text Available A series of metal cation modified layered-double hydroxides (LDHs and mixed oxides were prepared and used to be the selective oxidation of toluene with O2. The results revealed that the modified LDHs exhibited much higher catalytic performance than their parent LDH and the modified mixed oxides. Moreover, the metal cations were also found to play important roles in the catalytic performance and stabilities of modified catalysts. Under the optimal reaction conditions, the highest toluene conversion reached 8.7% with 97.5% of the selectivity to benzyldehyde; moreover, the catalytic performance remained after nine catalytic runs. In addition, the reaction probably involved a free-radical mechanism.

  16. 3D carbon/cobalt-nickel mixed-oxide hybrid nanostructured arrays for asymmetric supercapacitors.

    Science.gov (United States)

    Zhu, Jianhui; Jiang, Jian; Sun, Zhipeng; Luo, Jingshan; Fan, Zhanxi; Huang, Xintang; Zhang, Hua; Yu, Ting

    2014-07-23

    The electrochemical performance of supercapacitors relies not only on the exploitation of high-capacity active materials, but also on the rational design of superior electrode architectures. Herein, a novel supercapacitor electrode comprising 3D hierarchical mixed-oxide nanostructured arrays (NAs) of C/CoNi3 O4 is reported. The network-like C/CoNi3 O4 NAs exhibit a relatively high specific surface area; it is fabricated from ultra-robust Co-Ni hydroxide carbonate precursors through glucose-coating and calcination processes. Thanks to their interconnected three-dimensionally arrayed architecture and mesoporous nature, the C/CoNi3 O4 NA electrode exhibits a large specific capacitance of 1299 F/g and a superior rate performance, demonstrating 78% capacity retention even when the discharge current jumps by 100 times. An optimized asymmetric supercapacitor with the C/CoNi3 O4 NAs as the positive electrode is fabricated. This asymmetric supercapacitor can reversibly cycle at a high potential of 1.8 V, showing excellent cycling durability and also enabling a remarkable power density of ∼13 kW/kg with a high energy density of ∼19.2 W·h/kg. Two such supercapacitors linked in series can simultaneously power four distinct light-emitting diode indicators; they can also drive the motor of remote-controlled model planes. This work not only presents the potential of C/CoNi3 O4 NAs in thin-film supercapacitor applications, but it also demonstrates the superiority of electrodes with such a 3D hierarchical architecture. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. Calcium-based mixed oxide catalysts for methanolysis of Jatropha curcas oil to biodiesel

    International Nuclear Information System (INIS)

    Taufiq-Yap, Y.H.; Lee, H.V.; Hussein, M.Z.; Yunus, R.

    2011-01-01

    Calcium-based mixed oxides catalysts (CaMgO and CaZnO) have been investigated for the transesterification of Jatropha curcas oil (JCO) with methanol, in order to evaluate their potential as heterogeneous catalysts for biodiesel production. Both CaMgO and CaZnO catalysts were prepared by coprecipitation method of the corresponding mixed metal nitrate solution in the presence of a soluble carbonate salt at ∼ pH 8-9. The catalysts were characterized by X-ray diffraction (XRD), temperature programmed desorption of CO 2 (CO 2 -TPD), scanning electron microscopy (SEM) and N 2 adsorption (BET). The conversion of JCO by CaMgO and CaZnO were studied and compared with calcium oxide (CaO), magnesium oxide (MgO) and zinc oxide (ZnO) catalysts. Both CaMgO and CaZnO catalysts showed high activity as CaO and were easily separated from the product. CaMgO was found more active than CaZnO in the transesterification of JCO with methanol. Under the suitable transesterification conditions at 338 K (catalyst amount = 4 wt. %, methanol/oil molar ratio = 15, reaction time = 6 h), the JCO conversion of more than 80% can be achieved over CaMgO and CaZnO catalysts. Even though CaO gave the highest activity, the conversion of JCO decreased significantly after reused for forth run whereas the conversion was only slightly lowered for CaMgO and CaZnO after sixth run.

  18. Synthesis, characterization and thermal expansion studies on thorium-praseodymium mixed oxide solid solutions

    International Nuclear Information System (INIS)

    Panneerselvam, G.; Antony, M.P.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2010-01-01

    Full text: Thorium-praseodymium mixed oxide solid solutions containing 15, 25, 40 and 55 mole percent of praseodymia were synthesized by mixing the solutions of thorium nitrate in water and praseodymium oxide (Pr 6 O 11 ) in conc. HNO 3 . Subsequently, their hydroxides were co-precipitated by the addition of aqueous ammonia. Further the precipitate was dried at 50 deg C, calcined at 600 deg C for 4 hours and sintered at 1200 deg C for 6 h in air. X-ray diffraction measurements were performed for phase identification and lattice parameter derivation. Single-phase fluorite structure was observed for all the compositions. Bulk and theoretical densities of solid solutions were also determined by immersion and X-ray techniques. Thermal expansion coefficients and percentage linear thermal expansion of the solid solutions were determined using high temperature X-ray diffraction technique in the temperature range 300 to 1700 K for the first time. The room temperature lattice constants estimated for above compositions are 0.5578, 0.5565, 0.5545 and 0.5526 nm, respectively. The mean linear thermal expansion coefficients for the solid solutions are 15.48 x 10 -6 K -1 , 18.35 x 10 -6 K -1 , 22.65 x 10 -6 K -1 and 26.95 x 10 -6 K -1 , respectively. The percentage linear thermal expansions in this temperature range are 1.68, 1.89, 2.21 and 2.51 respectively. It is seen that the solid solutions are stable up to 1700 K. It is also seen that the effect and nature of the dopant are the important parameters influencing the thermal expansion of the ThO 2 . The lattice parameter of the solid solutions exhibited a decreasing trend with respect to praseodymia addition. The percentage linear thermal expansion of the solid solutions increases steadily with increasing temperature

  19. Water reactivity with mixed oxide (U,Pu)O2 surfaces

    International Nuclear Information System (INIS)

    Gaillard, Jeremy

    2013-01-01

    The interaction of water with actinides oxide surfaces remains poorly understood. The adsorption of water on PuO 2 surface and (U,Pu)O 2 surface leads to hydrogen generation through radiolysis but also surface evolution. The study of water interaction with mixed oxide (U,Pu)O 2 and PuO 2 surfaces requires the implementation of non intrusive techniques. The study of the hydration of CeO 2 surface is used to study the effectiveness of different techniques. The results show that the water adsorption leads to the surface evolution through the formation of a hydroxide superficial layer. The reactivity of water on the surface depends on the calcination temperature of the oxide precursor. The thermal treatment of hydrated surfaces can regenerate the surface. The study on CeO 2 hydration emphasizes the relevancies of these techniques in studying the hydration of surfaces. The hydrogen generation through water radiolysis is studied with an experimental methodology based on constant relative humidity in the radiolysis cell. The hydrogen accumulation is linear for the first hours and then tends to a steady state content. A mechanism of hydrogen consumption is proposed to explain the existence of the steady state of hydrogen content. This mechanism enables to explain also the evolution of the oxide surface during hydrogen generation experiments as shown by the evolution of hydrogen accumulation kinetics. The accumulation kinetics depends on the dose rate, specific surface area and the relative humidity but also on the oxide aging. The plutonium percentage appears to be a crucial parameter in hydrogen accumulation kinetics. (author) [fr

  20. Medium-Index Mixed-Oxide Layers for Use in AR-Coatings

    Science.gov (United States)

    Ganner, Peter

    1986-10-01

    Ttedesign philosophy of MC-AR-Coatings can be divided into two categories: a) Restriction to two film materials, namely one high-index and one low-index material and b) Use of medium-index layers in addition to high- and low-index layers. Both philosophies have advan-tages and drawbacks. In case a) the total number of layers necessary to obtain a required reflectance curve has to be higher. Thus in case of production errors it can be a problem to find out which layer was responsible for a deviation of the measured reflectance from the nominal one. In case b) using more than two materials reduces the total number of layers and consequently, pinpointing the cause of even small production errors is made simpler. Unfortunately there are not many materials commercially available which can be used to make hard, durable and robust films in the medium-index range namely between n=1.65 and n=2.00. In this paper the results of homogeneous mixtures of Alumina (Al203) and Tantala (Ta205) used for EB-gun evaporated medium-index films in AR-coatings is presented. It is shown that by proper adjustment of the weight percentages of the oxide mixture one can get homogeneous films in this index range. A number of design examples show the favourable application of such layers in AR-coatings. Among the most important ones is the well known QHQ-design for BBAR-coatings as well as AR-designs of the multiple half wave type with extended bandwidth. Further applications of the mixed-oxide layers are AR-coatings for cemented optical elements and beam splitters.

  1. Modification of Co-Mn-Al Mixed Oxide with Potassium and Its Effect on Deep Oxidation of VOC

    Czech Academy of Sciences Publication Activity Database

    Jirátová, Květa; Mikulová, Jana; Klempa, Jan; Grygar, Tomáš; Bastl, Zdeněk; Kovanda, F.

    2009-01-01

    Roč. 361, 1-2 (2009), s. 106-116 ISSN 0926-860X R&D Projects: GA ČR GA104/07/1400 Institutional research plan: CEZ:AV0Z40720504; CEZ:AV0Z40320502; CEZ:AV0Z40400503 Keywords : mixed oxide catalysts * VOC total oxidation * potassium promoter Subject RIV: CC - Organic Chemistry Impact factor: 3.564, year: 2009

  2. Supported Layered Double Hydroxide-Related Mixed Oxides and Their Application in the Total Oxidation of Volatile Organic Compounds

    Czech Academy of Sciences Publication Activity Database

    Kovanda, F.; Jirátová, Květa

    2011-01-01

    Roč. 53, č. 2 (2011), s. 305-316 ISSN 0169-1317 R&D Projects: GA ČR GAP106/10/1762; GA ČR GA106/09/1664 Institutional research plan: CEZ:AV0Z40720504 Keywords : layered double hydroxides * hydrothermal reaction * mixed oxides Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.474, year: 2011

  3. Optimization of Cs Content in Co-Mn-Al Mixed Oxide as Catalyst for N2O Decomposition.

    Czech Academy of Sciences Publication Activity Database

    Chromčáková, Ž.; Obalová, L.; Kustrowski, P.; Drozdek, M.; Karásková, K.; Jirátová, Květa; Kovanda, F.

    2015-01-01

    Roč. 41, č. 12 (2015), s. 9319-9332 ISSN 0922-6168. [Pannonian Symposium on Catalysis /12./. Třešť, 16.09.2014-20.09.2014] R&D Projects: GA ČR GA14-13750S Institutional support: RVO:67985858 Keywords : nitrous oxide * mixed oxide catalysts * cesium promoter * layered double hydroxides Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 1.833, year: 2015

  4. The behavior of SiC and Si3N4 ceramics in mixed oxidation/chlorination environments

    Science.gov (United States)

    Marra, John E.; Kreidler, Eric R.; Jacobson, Nathan S.; Fox, Dennis S.

    1989-01-01

    The behavior of silicon-based ceramics in mixed oxidation/chlorination environments was studied. High pressure mass spectrometry was used to quantitatively identify the reaction products. The quantitative identification of the corrosion products was coupled with thermogravimetric analysis and thermodynamic equilibrium calculations run under similar conditions in order to deduce the mechanism of corrosion. Variations in the behavior of the different silicon-based materials are discussed. Direct evidence of the existence of silicon oxychloride compounds is presented.

  5. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  6. About preparation and properties of UC based fuel materials

    International Nuclear Information System (INIS)

    Vooght, D. de; Timmermans, W.; Batist, R. de.

    1978-07-01

    The sintering behaviour and the effect of a numer of production parameters on the properties of sintered UC materials have been studied. Materials investigated include slightly hyperstoichiometric UC(UCsub(1+x)), oxygen containing UC[U(CO)] and UC containing both oxygen and nitrogen [U(CON)]. The materials have been characterized in terms of grain size distribution for the pre-sintering powder, of porosity distribution for the powdered material and for the green and sintered pellets and of the density of the green and sintered pellets. Carbothermic reaction temperature, milling time, and to some extent sintering temperature have been varied. The report discusses the possible correlations between several parameters such as milling time, powder fineness, density, grain size of the sintered product, composition (O,N content), etc. (author)

  7. Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials

    International Nuclear Information System (INIS)

    Becker, Udo

    2013-01-01

    Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO 2 and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for 237 Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO 2 , and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity

  8. Effects of ethylene glycol ethers on diesel fuel properties and emissions in a diesel engine

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Cuenca, F.; Gomez-Marin, M. [Compania Logistica de Hidrocarburos (CLH), Central Laboratory, Mendez Alvaro 44, 28045 Madrid (Spain); Folgueras-Diaz, M.B., E-mail: belenfd@uniovi.es [Department of Energy, University of Oviedo, Independencia 13, 33004 Oviedo (Spain)

    2011-08-15

    Highlights: {yields} Effect of ethylene glycol ethers on diesel fuel properties. {yields} Effect of ethylene glycol ethers on diesel engine specific consumption and emissions. {yields} Blends with {<=}4 wt.% of oxygen do not change substantially diesel fuel quality. {yields} Blends with 1 and 2.5 wt.% of oxygen reduce CO and HC emissions, but not smoke. - Abstract: The effect of ethylene glycol ethers on both the diesel fuel characteristics and the exhaust emissions (CO, NO{sub x}, smoke and hydrocarbons) from a diesel engine was studied. The ethers used were monoethylene glycol ethyl ether (EGEE), monoethylene glycol butyl ether (EGBE), diethylene glycol ethyl ether (DEGEE). The above effect was studied in two forms: first by determining the modification of base diesel fuel properties by using blends with oxygen concentration around 4 wt.%, and second by determining the emission reductions for blends with low oxygen content (1 wt.%) and with 2.5 wt.% of oxygen content. The addition of DEGEE enhances base diesel fuel cetane number, but EGEE and EGBE decrease it. For concentrations of {>=}4 wt.% of oxygen, EGEE and diesel fuel can show immiscibility problems at low temperatures ({<=}0 {sup o}C). Also, every oxygenated compound, according to its boiling point, modifies the distillation curve at low temperatures and the distillate percentage increases. These compounds have a positive effect on diesel fuel lubricity, and slightly decrease its viscosity. Blends with 1 and 2.5 wt.% oxygen concentrations were used in order to determine their influence on emissions at both full and medium loads and different engine speeds. Generally, all compounds help to reduce CO, and hydrocarbon emissions, but not smoke. The best results were obtained for blends with 2.5 wt.% of oxygen. At this concentration, the additive efficiency in decreasing order was EGEE > DEGEE > EGBE for CO emissions and DGEE > EGEE > EGBE for hydrocarbon emissions. For NO{sub x}, both its behaviour and the

  9. MOX fuel for Indian nuclear power programme

    International Nuclear Information System (INIS)

    Kamath, H.S.; Anantharaman, K.; Purushotham, D.S.C.

    2000-01-01

    A sound energy policy and a sound environmental policy calls for utilisation of plutonium (Pu) in nuclear power reactors. The paper discusses the use of Pu in the form of mixed oxide (MOX) fuel in two Indian boiling water reactors (BWRs) at Tarapur. An industrial scale MOX fuel fabrication plant is presently operational at Tarapur which is capable of manufacturing MOX fuels for BWRs and in future for PHWRs. The plant can also manufacture mixed oxide fuel for prototype fast breeder reactor (PFBR) and development work in this regard has already started. The paper describes the MOX fuel manufacturing technology and quality control techniques presently in use at the plant. The irradiation experience of the lead MOX assemblies in BWRs is also briefly discussed. The key areas of interest for future developments in MOX fuel fabrication technology and Pu utilisation are identified. (author)

  10. Testing of wet scrap recovery equipment for mixed oxide scrap reprocessing

    International Nuclear Information System (INIS)

    Demiter, J.A.; Klem, M.J.; Owen, T.J.

    1984-08-01

    The Wet Scrap Recovery (WSR) program was initiated at the Hanford Engineering Development Laboratory (HEDL) by Westinghouse Hanford Company in Richland, Washington to demonstrate fuel fabrication scrap recovery and reconversion to fuel grade oxide powder using the continuous coprecipitation-calcination (COPRECAL) conversion process. Advancements in process control equipment and instrumentation were also developed and demonstrated

  11. Handbook on process and chemistry of nuclear fuel reprocessing. 3rd edition

    International Nuclear Information System (INIS)

    2015-03-01

    The fundamental data on spent nuclear fuel reprocessing and related chemistry was collected and summarized as a new edition of 'Handbook on Process and Chemistry of Nuclear Fuel Reprocessing'. The purpose of this handbook is contribution to development of the fuel reprocessing and fuel cycle technology for uranium fuel and mixed oxide fuel utilization. Contents in this book was discussed and reviewed by specialists of science and technology on fuel reprocessing in Japan. (author)

  12. Effect of solid fission products forming dissolved oxide(Nd) and metallic precipitate(Ru) on the thermophysical properties of MOX fuel

    International Nuclear Information System (INIS)

    Kim, Dong Joo

    2006-02-01

    This study experimentally investigated the effect of solid fission products on the thermophysical properties of the mixed oxide fuel and evaluated them on the basis of the analytical theory. Neodymium and ruthenium were selected for the experiments to represent the physical states of the solid fission product as a 'dissolved oxide' and 'metallic precipitate', respectively. The state of the additives, crystal structures, lattice parameters, and theoretical densities were investigated with X-ray diffraction (XRD). Thermal diffusivities and thermal expansion rates were measured with laser flash method and dilatometry, respectively. The thermal expansion data were then fitted to obtain an correlation equation of the density variation as a function of the temperature. The specific heat capacity values were determined using the Neumann-Kopp's rule. The thermal expansion of the 'Nd.added' sample linearly increased with the concentration of the neodymium, which is primarily due to the fact that the melting point of Nd 2 O 3 is lower than that of UO 2 . On the other hand, the thermal expansion of the 'Ru.added' sample hardly changed with increasing ruthenium content. Thermal conductivities of the simulated MOX fuel were determined on the basis of the thermal diffusivities, density variation, and specific heat values measured in this study. The effect of additives on the thermal conductivity of the samples was quantified in the form of the thermal resistance equation, the reciprocal of the phonon conduction equation, which was determined from measured data. For 'dissolved oxide' sample in the UO 2 matrix, the effect is mainly attributed to the increase of lattice point defects caused by U 4+ , Ce 4+ , Nd 3+ and O 2- ions, which play the role of phonon scattering centers, that is, mean free path of phonon scattering decreases with the point defects, thus increase the thermal resistance. Also, the mass difference between the host (U) and the substituted atom (Ce and/or Nd) can

  13. Impact of fuel properties on advanced power systems

    Energy Technology Data Exchange (ETDEWEB)

    Sondreal, E.A.; Jones, M.L.; Hurley, J.P.; Benson, S.A.; Willson, W.G. [Univ. of North Dakota, Grand Forks, ND (United States)

    1995-12-01

    Advanced coal-fired combined-cycle power systems currently in development and demonstration have the goal of increasing generating efficiency to a level approaching 50% while reducing the cost of electricity from new plants by 20% and meeting stringent standards on emissions of SO{sub x} NO{sub x} fine particulates, and air toxic metals. Achieving these benefits requires that clean hot gas be delivered to a gas turbine at a temperature approaching 1350{degrees}C, while minimizing energy losses in the gasification, combustion, heat transfer, and/or gas cleaning equipment used to generate the hot gas. Minimizing capital cost also requires that the different stages of the system be integrated as simply and compactly as possible. Second-generation technologies including integrated gasification combined cycle (IGCC), pressurized fluidized-bed combustion (PFBC), externally fired combined cycle (EFCC), and other advanced combustion systems rely on different high-temperature combinations of heat exchange, gas filtration, and sulfur capture to meet these requirements. This paper describes the various properties of lignite and brown coals.

  14. Fabrication and properties of carbon network reinforced composite fuel

    International Nuclear Information System (INIS)

    Umer, Malik Adeel; Mistarihi, Qusai Mahmoud; Kim, Joon Hui; Hong, Soon Hyung; Ryu, Ho Jin

    2014-01-01

    Zirconium dioxide composites reinforced with 3D glassy carbon foam was fabricated using Spark Plasma Sintering (SPS) with a heating rate of 100degC/min and a uniaxial pressure of 50 MPa at 1500degC, 1600degC, and 1700degC, respectively. The effect of carbon foam on the thermal properties of the ZrO 2 composites was investigated. In addition, the effect of the sintering temperature on the densification of the composites was also investigated and the optimized sintering temperature was identified. The microstructures of 3D carbon foam reinforced ZrO 2 composites showed that the 3D shape of carbon foam was retained after the sintering process, and the ZrO 2 was homogeneously distributed within the 3D carbon foam. At the interfaces between the 3D carbon foam and ZrO 2 , neither a chemical reaction nor a new phase formation was detected by Scanning Electron Microscopy (SEM) and X-ray Diffractometry (XRD). The thermal diffusivity of carbon foam reinforced ZrO 2 composites measured at 1100degC was increased by 47% and reached to 0.66 mm 2 s -1 and the thermal conductivity was increased by 50% and reached to 2.428 W/m-K. (author)

  15. Improvement of the stability of basic mixed oxides used as catalysts for aldol condensation of bio-derived compounds by palladium addition

    International Nuclear Information System (INIS)

    Faba, Laura; Díaz, Eva; Ordóñez, Salvador

    2013-01-01

    Aqueous-phase aldol condensation of biomass-derived ketones and aldehydes is a key step in the preparation of fuels and chemicals from renewable resources. Furfural–acetone aldol condensation yielding C 8 and C 13 adducts was studied at 323 K and 1 MPa in a stirred batch reactor. We propose a new strategy for minimizing catalytic deactivation, consisting of modifying catalysts (MgO–ZrO 2 and MgO–Al 2 O 3 ) by Pd addition (2%). This modification slightly changes the morphology and surface chemistry of the supports, leading to changes on the catalysts performance but not reaction mechanism modifications. If condensation is performed in hydrogen atmosphere, the partial hydrogenation of the condensation adducts increases its water solubility, minimizing catalyst deactivation. In that way, the selectivity for C13 adduct decreases only 25% between two successive reaction cycles using Pd/MgO–ZrO 2 catalyst, whereas this decrease is of 90% for the un-doped mixed oxides. These effects are less marked for the MgO–Al 2 O 3 catalyst. Highlights: •Furfural and acetone aldolization on Pd/MgO–ZrO 2 and Pd/MgO–Al 2 O 3 . •Conversion and selectivity towards C8 and C13 low and unstable catalysts. •Aldolization in H 2 decreases coke formation: higher reusability of the catalysts

  16. Fuel cycle parameters for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-05-01

    This report summarizes seven fuel cycle parameters (efficiency, specific power, burnup, equilibrium net fissile feed, equilibrium net fissile surplus, first charge fissile content, and whether or not fuel reprocessing is required) to be used in long-term strategy analyses of fuel cycles based on natural UO 2 , low enriched uranium, mixed oxides, plutonium topped thorium, uranium topped thorium, and the fast breeder oxide cycle. (LL)

  17. Tuning the acidity of niobia: Characterization and catalytic activity of Nb{sub 2}O{sub 5}–MeO{sub 2} (Me = Ti, Zr, Ce) mesoporous mixed oxides

    Energy Technology Data Exchange (ETDEWEB)

    Stošić, Dušan; Bennici, Simona [Université Lyon 1, CNRS, UMR 5256, IRCELYON, Institut de recherches sur la catalyse et l' environnement de Lyon, 2 avenue Albert Einstein, F-69626 Villeurbanne (France); Pavlović, Vladimir; Rakić, Vesna [Faculty of Agriculture, Department of Chemistry, University of Belgrade, Nemanjina 6, 11080 Zemun (Serbia); Auroux, Aline, E-mail: aline.auroux@ircelyon.univ-lyon1.fr [Université Lyon 1, CNRS, UMR 5256, IRCELYON, Institut de recherches sur la catalyse et l' environnement de Lyon, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)

    2014-08-01

    Mesoporous Nb{sub 2}O{sub 5}–MeO{sub 2} (Me = Ti, Zr, Ce) mixed oxides were successfully prepared using evaporation-induced self-assembly (EISA) method. The structural and textural properties of these materials have been fully characterized using appropriate techniques (low-temperature adsorption–desorption of nitrogen, thermogravimetric analysis, X-ray diffraction analysis (XRD) transmission electron microscopy (TEM), scanning electron microscopy (SEM) and Raman spectroscopy). Acid–base properties were estimated by adsorption microcalorimetry of NH{sub 3} and SO{sub 2} molecules in order to determine the population, strength and strength distribution of acidic or basic sites. Formation of mesoporous structure was confirmed by the results of XRD, TEM and BET techniques. Results of adsorption microcalorimetry technique showed that the type of transition metal oxide added to niobia has a decisive role for acidic-basic character of investigated mixed oxides. Among the investigated mixed oxide formulations only Nb{sub 2}O{sub 5}–CeO{sub 2} was amphoteric, while the other samples showed prominent acidic character. All the investigated materials are catalytically active in fructose dehydration; conversion of fructose and selectivity to 5-hydroxymethylfurfural (5-HMF) and levulinic acid (LA) are proved to be dependant on the number of acidic sites on the surface of catalysts. Furthermore, presence of the basic sites on the surface of the catalyst decreases the activity in the fructose dehydration reaction, as in the case of Nb{sub 2}O{sub 5}–CeO{sub 2} sample. - Highlights: • Mesoporous Nb{sub 2}O{sub 5}–MeO{sub 2} mixed oxides were successfully prepared by EISA method. • Acidic–basic properties depend on the nature of the oxide that was mixed with niobia. • Catalytic activity was tested in fructose dehydration in aqueous phase. • Selectivity and conversion in reaction are correlated to the number of acid sites.

  18. Experimental verification of the thermodynamic properties for a jet-A fuel

    Science.gov (United States)

    Graciasalcedo, Carmen M.; Brabbs, Theodore A.; Mcbride, Bonnie J.

    1988-01-01

    Thermodynamic properties for a Jet-A fuel were determined by Shell Development Company in 1970 under a contract for NASA Lewis Research Center. The polynomial fit necessary to include Jet-A fuel (liquid and gaseous phases) in the library of thermodynamic properties of the NASA Lewis Chemical Equilibrium Program is calculated. To verify the thermodynamic data, the temperatures of mixtures of liquid Jet-A injected into a hot nitrogen stream were experimentally measured and compared to those calculated by the program. Iso-octane, a fuel for which the thermodynamic properties are well known, was used as a standard to calibrate the apparatus. The measured temperatures for the iso-octane/nitrogen mixtures reproduced the calculated temperatures except for a small loss due to the non-adiabatic behavior of the apparatus. The measurements for Jet-A were corrected for this heat loss and showed excellent agreement with the calculated temperatures. These experiments show that this process can be adequately described by the thermodynamic properties fitted for the Chemical Equilibrium Program.

  19. Mechanical Properties and Structures of Pyrolytic Carbon Coating Layer in HTR Coated Particle Fuel

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Young Min; Kim, Woong Ki; Cho, Moon Sung

    2009-01-01

    The TRISO(tri-isotropic)-coated fuel particle for a HTR(High Temperature gas-cooled Reactor) has a diameter of about 1 mm, composed of a nuclear fuel kernel and four different outer coating layers, consisting of a buffer PyC (pyrolytic carbon) layer, inner PyC layer, SiC layer, and outer PyC layer with different coating thicknesses following a specific fuel design. While the fuel kernel is a source for a heat generation by a nuclear fission of fissile uranium, each of the four coating layers acts as a different role in view of retaining the generated fission products and the other interactions during an in-reactor service. Among these coating layers, PyC properties are scarcely in agreement among various investigators and the dependency of their changes upon the deposition condition is comparatively large due to their additional anisotropic properties. Although a recent review work has contributed to an establishment of relationship between the material properties and QC measurements, the data on the mechanical properties and structural parameters of PyC coating layers remain still unclearly evaluated. A review work on dimensional changes of PyC by neutron irradiation was one of re-evaluative works recently attempted by the authors. In this work, an attempt was made to analyze and re-evaluate the existing data of the experimental results of the mechanical properties, i.e., Young's modulus and fracture stress, in relation with the coating conditions, density and the BAF (Bacon Anisotropy Factor), an important structural parameter, of PyC coating layers obtained from various experiments performed in the early periods of the HTR coated particle development

  20. The influence of the mechanical properties on fuel rod support characteristics – A case study of dual cooled fuel

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Kim, Hyung Kyu; Ko, Sung Ho

    2014-01-01

    Highlights: • Spring characterization test and analysis were performed to obtain characteristic curves of modified H-type spring. • Using an actual mechanical property is needed to correctly predict the spring characteristics. • The characteristics during unloading should be used for a spacer grid support design. - Abstract: This paper concerns a finite element analysis for a spacer grid support (spring and dimple) design. An accurate prediction of the support characteristics (contact force vs. deflection) is the most crucial in the design by analysis. It is found that the mechanical properties are the key parameter to simulate the characteristics as close as the experimental results after using three different sets of mechanical property data including the actual tensile test results of the present material for a spacer grid of a dual cooled fuel. Besides, the validity of using the characteristics during unloading process is also discussed incorporating a possible overshoot of the support. The coincidence between the present finite element prediction and experimental results is quite good: less than 3.09% at most