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Sample records for fuel dissolver solution

  1. Dissolution of powdered spent fuel and U crystallization from actual dissolver solution for 'NEXT' process development

    International Nuclear Information System (INIS)

    Nomura, Kazunori; Hinai, Hiroshi; Nakahara, Masaumi; Kaji, Naoya; Kamiya, Masayoshi; Ohyama, Koichi; Sano, Yuichi; Washiya, Tadahiro; Komaki, Jun

    2008-01-01

    The beaker-scale experiments on the effective powdered fuel dissolution and the U crystallization from dissolver solution with the irradiated MOX fuel from the experimental fast reactor 'JOYO' were carried out. The powdered fuel was effectively dissolved into the nitric acid solution. In the U crystallization experiments, U crystal was obtained from the actual dissolver solution without any addition of reagent. (authors)

  2. Selection of dissolution process for spent fuels and preparation of corrosion test solution simulated to dissolver (contract research)

    International Nuclear Information System (INIS)

    Motooka, Takafumi; Terakado, Shogo; Koya, Toshio; Hamada, Shozo; Kiuchi, Kiyoshi

    2001-03-01

    In order to evaluate the reliability of reprocessing equipment materials used in the Rokkasho Reprocessing Plant, we have proceeded a mock-up test and laboratory tests for getting corrosion parameters. In a dissolver made of zirconium, the simulation of test solutions to the practical solution which includes the high concentration of radioactive elements such as FP and TRU is one of the important issues with respect to the life prediction. On this experiment, the dissolution process of spent fuels and the preparation of test solution for evaluating the corrosion resistance of dissolver materials were selected. These processes were tested in the No.3 cell of WASTEF. The test solution for corrosion tests was prepared by adjusting the uranium and nitric acid concentrations. (author)

  3. Resin bead-thermal ionization mass spectrometry for determination of plutonium concentration in irradiated fuel dissolver solution

    International Nuclear Information System (INIS)

    Paul, Sumana; Shah, R.V.; Aggarwal, S.K.; Pandey, A.K.

    2015-01-01

    Determination of isotopic composition (IC) and concentration of plutonium (Pu) is necessary at various stages of nuclear fuel cycle which involves analysis of complex matrices like dissolver solution of irradiated fuel, nuclear waste stream etc. Mass spectrometry, e.g. thermal ionization mass spectrometry (TIMS) and inductively coupled plasma mass spectrometry (ICP-MS) are commonly used for determination of IC and concentration of plutonium. However, to circumvent matrix interferences, efficient separation as well as preconcentration of Pu is required prior to mass spectrometric analysis. Purification steps employing ion-exchange resins are widely used for the separation of Pu from dissolver solution or from mixture of other actinides e.g. U, Am. However, an alternative way is to selectively preconcentrate Pu on a resin bead, followed by direct loading of the bead on the filament of thermal ionization mass spectrometer

  4. Dissolving method for nuclear fuel oxide

    International Nuclear Information System (INIS)

    Tomiyasu, Hiroshi; Kataoka, Makoto; Asano, Yuichiro; Hasegawa, Shin-ichi; Takashima, Yoichi; Ikeda, Yasuhisa.

    1996-01-01

    In a method of dissolving oxides of nuclear fuels in an aqueous acid solution, the oxides of the nuclear fuels are dissolved in a state where an oxidizing agent other than the acid is present together in the aqueous acid solution. If chlorate ions (ClO 3 - ) are present together in the aqueous acid solution, the chlorate ions act as a strong oxidizing agent and dissolve nuclear fuels such as UO 2 by oxidation. In addition, a Ce compound which generates Ce(IV) by oxidation is added to the aqueous acid solution, and an ozone (O 3 ) gas is blown thereto to dissolve the oxides of nuclear fuels. Further, the oxides of nuclear fuels are oxidized in a state where ClO 2 is present together in the aqueous acid solution to dissolve the oxides of nuclear fuels. Since oxides of the nuclear fuels are dissolved in a state where the oxidizing agent is present together as described above, the oxides of nuclear fuels can be dissolved even at a room temperature, thereby enabling to use a material such as polytetrafluoroethylene and to dissolve the oxides of nuclear fuels at a reduced cost for dissolution. (T.M.)

  5. Isotope dilution alpha spectrometry for the determination of plutonium concentration in irradiated fuel dissolver solution : IDAS and R-IDAS

    International Nuclear Information System (INIS)

    Ramaniah, M.V.; Jain, H.C.; Aggarwal, S.K.; Chitambar, S.A.; Kavimandan, V.D.; Almaula, A.I.; Shah, P.M.; Parab, A.R.; Sant, V.L.

    1980-01-01

    The report presents a new technique, Isotope Dilution Alpha Spectrometry (IDAS) and Reverse Isotope Dilution Alpha Spectrometry (R-IDAS) for determining the concentration of plutonium in the irradiated fuel dissolver solution. The method exploits sup(238)Pu in IDAS and sup(239)Pu in R-IDAS as a spike and provides an alternative method to Isotope Dilution Mass Spectrometry (IDMS) which requires enriched sup(242)Pu as a spike. Depending upon the burn-up of the fuel, sup(238)Pu or sup(239)Pu is used as a spike to change the sup(238)Pu/(sup(239)Pu+sup(240)Pu)α activity ratio in the sample by a factor of 10. This change is determined by α-spectrometry on electrodeposited sources using a solid state silicon surface barrier detector coupled to a multichannel analyser. The validity of a simple method based on the geometric progression (G.P.) decrease for the far tail of the spectrum to correct for the tail contribution of sup(238)Pu peak (5.50 MeV) to the low energy sup(239)Pu + sup(240)Pu peak (5.17 MeV) is established. Results for the plutonium concentration on different irradiated fuel dissolver solutions with burn-uo ranging from J,000 to 100,000 MWD/TU are presented and compared with those obtained by IDMS. The values obtained by IDAS or R-IDAS and IDMS agree within 0.5%. (auth.)

  6. An intercomparison experiment on isotope dilution thermal ionisation mass spectrometry using plutonium-239 spike for the determination of plutonium concentration in dissolver solution of irradiated fuel

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Shah, P.M.; Saxena, M.K.; Jain, H.C.; Gurba, P.B.; Babbar, R.K.; Udagatti, S.V.; Moorthy, A.D.; Singh, R.K.; Bajpai, D.D.

    1996-01-01

    Determination of plutonium concentration in the dissolver solution of irradiated fuel is one of the key measurements in the nuclear fuel cycle. This report presents the results of an intercomparison experiment performed between Fuel Chemistry Division (FCD) at BARC and PREFRE, Tarapur for determining plutonium concentration in dissolver solution of irradiated fuel using 239 Pu spike in isotope dilution thermal ionisation mass spectrometry (ID-TIMS). The 239 Pu spike method was previously established at FCD as viable alternative to the imported enriched 242 Pu or 244 Pu; the spike used internationally for plutonium concentration determination by IDMS in dissolver solution of irradiated fuel. Precision and accuracy achievable for determining plutonium concentration are compared under the laboratory and the plant conditions using 239 Pu spike in IDMS. For this purpose, two different dissolver solutions with 240 Pu/ 239 Pu atom ratios of about 0.3 and 0.07 corresponding, respectively, to high and low burn-up fuels, were used. The results of the intercomparison experiment demonstrate that there is no difference in the precision values obtained under the laboratory and the plant conditions; with mean precision values of better than 0.2%. Further, the plutonium concentration values determined by the two laboratories agreed within 0.3%. This exercise, therefore, demonstrates that ID-TIMS method using 239 Pu spike can be used for determining plutonium concentration in dissolver solution of irradiated fuel, under the plant conditions. 7 refs., 8 tabs

  7. Simultaneous determination of uranium and plutonium in dissolver solution of irradiated fuel, using ID-TIMS. IRP-11

    International Nuclear Information System (INIS)

    Shah, Raju; Sasi Bhushan, K.; Govindan, R.; Alamelu, D.; Khodade, P.S.; Aggarwal, S.K.

    2007-01-01

    A simple sample preparation and simultaneous analysis method to determine uranium and plutonium from dissolver solution, employing the technique of Isotope Dilution Mass spectrometry has been demonstrated. The method used, co-elusion of Uranium and Plutonium from anion exchanger column after initial elution of major part of uranium in 1:5 HNO 3 in order to reduce the initial U/Pu ratio from 1000 to about 100-200 in the co-eluted fraction. Due to the availability of variable multi-collector system, different Faraday cups were adjusted to collect the different ion intensities corresponding to the different masses, during the simultaneous analysis of Uranium and Plutonium, loaded on Re double filament assembly. 233 U and PR grade Plutonium were used as spikes to determine Uranium and Plutonium from dissolver solution of irradiated fuel from research reactor. The possibility of getting the isotopic composition of uranium from the simultaneous analysis of co-eluted purified fraction of U and Pu from spiked aliquots is also explained. (author)

  8. Radiometric determination of 90Sr in the dissolver solution of the spent PHWR fuel after its separation with solvent extraction and extraction chromatography

    International Nuclear Information System (INIS)

    Kulkarni, P.G.; Gupta, K.K.; Pant, D.K.; Bhalerao, B.A.; Gurba, P.B.; Janardan, P.; Changrani, R.D.; Dey, P.K.; Pathak, P.N.; Mohapatra, P.K.; Manchanda, V.K.

    2010-01-01

    A simple radiometric method for 90 Sr determination in the dissolver solution of the PHWR spent fuel has been developed.The method involves the quantitative separation of Sr from the associated actinides and other fission products by solvent extraction with 30% trialkylphosphine oxide (TRPO) -n-dodecane followed by extraction chromatography with XAD-7-Di-butylcyclohexano-18-crown-6 resin. The separation scheme yields quantitative recovery of 90 Sr and the separated 90 Sr was found to be radiochemically pure. 90 Sr was estimated by β-radiometry and the precision of the method at 5 mCi/mL level was 2% (RSD). (author)

  9. Clarification of dissolved irradiated light-water-reactor fuel

    International Nuclear Information System (INIS)

    Rodrigues, G.C.

    1983-02-01

    Bench-scale studies with actual dissolved irradiated light water reactor (LWR) fuels showed that continuous centrifugation is a practical clarification method for reprocessing. Dissolved irradiated LWR fuel was satisfactorily clarified in a bench-scale, continuous-flow bowl centrifuge. The solids separated were successfully reslurried in water. When the reslurried solids were mixed with clarified centrate, the resulting suspension behaved similar to the original dissolver solution during centrifugation. Settling rates for solids in actual irradiated fuel solutions were measured in a bottle centrifuge. The results indicate that dissolver solutions may be clarified under conditions achievable by available plant-scale centrifuge technology. The effective particle diameter of residual solids was calculated to be 0.064 microns for Oconee-1 fuel and 0.138 microns for Dresden-1 fuel. Filtration was shown unsuitable for clarification of LWR fuel solutions. Conventional filtration with filter aid would unacceptably complicate remote canyon operation and maintenance, might introduce dissolved silica from filter aids, and might irreversibly plug the filter with dissolver solids. Inertial filtration exhibited irreversible pluggage with nonradioactive stand-in suspensions under all conditions tested

  10. Development of a novel solvent for the simultaneous separation of strontium and cesium from dissolved Spent Nuclear Fuel solutions

    International Nuclear Information System (INIS)

    Catherine L. Riddle; John D. Baker; Jack D. Law; Christopher A. McGrath; David H. Meikrantz; Bruce J. Mincher; Dean R. Peterman; Terry A. Todd

    2004-01-01

    The recovery of Cs and Sr from acidic solutions by solvent extraction has been investigated. The goal of this project was to develop an extraction process to remove Cs and Sr from high-level waste in an effort to reduce the heat loading in storage. Solvents for the extraction of Cs and Sr separately have been used on both caustic and acidic spent nuclear fuel waste in the past. The objective of this research was to find a suitable solvent for the extraction of both Cs and Sr simultaneously from acidic nitrate media. The solvents selected for this research possess good stability and extraction behavior when mixed together. The extraction experiments were performed with 4,4,(5)-Di-(tbutyldicyclohexano)-18-crown-6 (DtBuCH18C6), Calix[4]arene-bis-(tert-octylbenzocrown-6) (BOBCalixC6) and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (Cs-7SB modifier) in a branched aliphatic kerosene (Isopar L). The BOBCalixC6 and Cs-7SB modifier were developed at Oak Ridge National Laboratory (ORNL) by Bonnesen et al. [1]. The values obtained from the SREX solvent for DSr in 1 M nitric acid ranged from 0.7 to 2.2 at 25 C and 10 C respectively. The values for DCs in 1 M nitric acid with the CSSX solvent ranged from 8.0 to 46.0 at 25 C and 10 C respectively. A new mixed solvent, developed at the Idaho National Engineering and Environmental Laboratory (INEEL) by Riddle et al. [2], showed distributions for Sr ranging from 8.8 to 17.4 in 1 M nitric acid at 25 C and 10 C respectively. The DCs for the mixed solvent ranged from 7.7 to 20.2 in 1 M nitric acid at 25 C to 10 C respectively. The unexpectedly high distributions for Sr at both 25 C and 10 C show a synergy in the mixed solvent. The DCs, although lower than with CSSX solvent, still showed good extraction behavior

  11. Characterizing Dissolved Gases in Cryogenic Liquid Fuels

    Science.gov (United States)

    Richardson, Ian A.

    Pressure-Density-Temperature-Composition (PrhoT-x) measurements of cryogenic fuel mixtures are a historical challenge due to the difficulties of maintaining cryogenic temperatures and precision isolation of a mixture sample. For decades NASA has used helium to pressurize liquid hydrogen propellant tanks to maintain tank pressure and reduce boil off. This process causes helium gas to dissolve into liquid hydrogen creating a cryogenic mixture with thermodynamic properties that vary from pure liquid hydrogen. This can lead to inefficiencies in fuel storage and instabilities in fluid flow. As NASA plans for longer missions to Mars and beyond, small inefficiencies such as dissolved helium in liquid propellant become significant. Traditional NASA models are unable to account for dissolved helium due to a lack of fundamental property measurements necessary for the development of a mixture Equation Of State (EOS). The first PrhoT-x measurements of helium-hydrogen mixtures using a retrofitted single-sinker densimeter, magnetic suspension microbalance, and calibrated gas chromatograph are presented in this research. These measurements were used to develop the first multi-phase EOS for helium-hydrogen mixtures which was implemented into NASA's Generalized Fluid System Simulation Program (GFSSP) to determine the significance of mixture non-idealities. It was revealed that having dissolved helium in the propellant does not have a significant effect on the tank pressurization rate but does affect the rate at which the propellant temperature rises. PrhoT-x measurements are conducted on methane-ethane mixtures with dissolved nitrogen gas to simulate the conditions of the hydrocarbon seas of Saturn's moon Titan. Titan is the only known celestial body in the solar system besides Earth with stable liquid seas accessible on the surface. The PrhoT-x measurements are used to develop solubility models to aid in the design of the Titan Submarine. NASA is currently designing the submarine

  12. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

  13. Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

    International Nuclear Information System (INIS)

    Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

    2011-01-01

    Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

  14. Advanced immobilization processes for fuel hulls and dissolver residues

    International Nuclear Information System (INIS)

    Hebel, W.; Boehme, G.; Findlay, J.R.; Sombret, C.

    1984-08-01

    Various research and development projects for the conditioning of cladding scraps and dissolver residues are pursued within the scope of the R and D programme on nuclear waste Management of the European Community. They include the characterization of the waste materials arising from industrial fuel reprocessing and the development of different waste immobilization techniques. These concern the embedment of scraps and residues into inert matrices like cement, metal alloys, compacted graphite and sintered ceramics as well as the treatment of the fuel hulls by melting or chemical conversion. The conditioned waste forms are tested as to their relevant properties for activity enclosure

  15. Advanced immobilization processes for fuel hulls and dissolver residues

    International Nuclear Information System (INIS)

    Hebel, W.; Boehme, G.; Findlay, J.R.; Sombert, C.

    1984-01-01

    Various research and development projects for the conditioning of cladding scraps and dissolver residues are pursued within the scope of the R and D programme on nuclear waste Management of the European Community. They include the characterization of the waste materials arising from industrial fuel reprocessing and the development of different waste immobilization techniques. These concern the embedment of the scraps and residues into inert matrices like cement, metal alloys, compacted graphite and sintered ceramics as well as the treatment of the fuel hulls by melting or chemical conversion. The conditioned waste forms are tested as to their relevant properties for activity enclosure

  16. Impact of solute concentration on the electrocatalytic conversion of dissolved gases in buffered solutions

    KAUST Repository

    Shinagawa, Tatsuya; Takanabe, Kazuhiro

    2015-01-01

    . These alterations of the electrolyte properties associated with the solute concentration are universally applicable to other aqueous gas-related electrochemical reactions because the currents are purely determined by mass transfer of the dissolved gases. © 2015

  17. Studies in the dissolver off-gas system for a spent FBR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Heinrich, E.; Huefner, R.; Weirich, F.

    1982-01-01

    Investigations of possible modifications of the process steps of a dissolver off-gas (DOG) system for a spent FBR fuel reprocessing plant are reported. The following operations are discussed: iodine removal from the fuel solution; behaviour of NOsub(x) and iodine in nitric acid off-gas scrubbers at different temperatures and nitric acid concentrations; iodine desorption from the scrub acid; selective absorption of noble gases in refrigerant-12; cold traps. The combination of suitable procedures to produce a total DOG system is described. (U.K.)

  18. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

  19. Nuclear safety of the ten-well insert for the SRP fuel element dissolver

    International Nuclear Information System (INIS)

    Perkins, W.C.; Forstner, J.L.

    1977-06-01

    Mass limits are developed and presented for safe dissolution of fissile materials in the Ten-Well Insert, an improved device for limiting the configuration of fuel in SRP dissolvers. This insert permits high-capacity dissolution of SRP fuels, offsite fuels, and scrap fissile materials with adequate margins of nuclear safety. Limits were developed by calculating the safe (subcritical) mass per well as a function of the concentration of fissile material in the dissolver solution. Safe mass values were then selected for use as well-loading limits so as to ensure subcriticality throughout the dissolution. Well-loading limits are presented for uranium metal, uranium-aluminum alloy, U 3 O 8 -aluminum cermet, plutonium-aluminum alloy, and uranium-plutonium-aluminum alloy. With these limits, the maximum k/sub eff/ is 0.95. Nuclear safety is maintained in process operations by conforming to well-loading limits calculated from the safe mass values, conforming to dissolver-loading limits, and maintaining the concentration of fissile material in solution below 4.0 g/l. 9 figures, 14 tables

  20. Specific application of burnup credit for MOX PWR fuels in the rotary dissolver

    International Nuclear Information System (INIS)

    Caplin, Gregory; Coulaud, Alexandre; Klenov, Pavel; Toubon, Herve

    2003-01-01

    In prospect of a Mixed OXide spent fuels processing in the rotary dissolver in COGEMA/La Hague plant, it is interesting to quantify the criticality-safety margins from the burnup credit. Using the current production computer codes and considering a minimal fuel irradiation of 3 200 megawatt-day per ton, this paper shows the impact of burnup credit on industrial parameters such as the permissible concentration in the dissolution solution or the permissible oxide mass in the rotary dissolver. Moreover, the burnup credit is broken down into five sequences in order to quantify the contribution of fissile nuclides decrease and of minor actinides and fission products formation. The implementation of the burnup credit in the criticality-safety analysis of the rotary dissolver may lead to workable industrial conditions for the particular MOX fuel studied. It can eventually be noticed that minor actinides contribution is negligible and that considering only the six major fission products is sufficient, owing to the weak fuel irradiation contemplated. (author)

  1. Fuel cell vehicles: technological solution

    International Nuclear Information System (INIS)

    Lopez Martinez, J. M.

    2004-01-01

    Recently it takes a serious look at fuel cell vehicles, a leading candidate for next-generation vehicle propulsion systems. The green house effect and air quality are pressing to the designers of internal combustion engine vehicles, owing to the manufacturers to find out technological solutions in order to increase the efficiency and reduce emissions from the vehicles. On the other hand, energy source used by currently propulsion systems is not renewable, the well are limited and produce CO 2 as a product from the combustion process. In that situation, why fuel cell is an alternative of internal combustion engine?

  2. Methods of conditioning waste fuel decladding hulls and dissolver residues

    International Nuclear Information System (INIS)

    De Regge, P.; Loida, A.; Schmidt-Hansberg, T.; Sombret, C.

    1985-01-01

    Several methods for conditioning spent fuel decladding hulls or dissolver residues have been considered in various countries of the European Community. Five of these methods used embedding technique with or without prior compaction: they are based on incorporation in metallic alloys, glass, ceramics, cements and metals or graphite compounds. A sixth one consists in melting the decladding materials. The corresponding research programmes have been pursued to varying states of progress with regard to demonstrating their feasibility on an industrial scale and the use of genuine wastes in bench scale experiments. The properties of the conditioned wastes have been investigated. Special attention has been paid to the corrosion resistance to various aqueous media as tap water, brine or clayey water. Although no categorical conclusion can be drawn from the initial results, the available findings provide a basis for assessing the different processes

  3. Some ideas about the modeling of experimental data obtained during spent fuel leaching in the presence of dissolved hydrogen

    International Nuclear Information System (INIS)

    Spahiu, K.

    2003-01-01

    Lately several experimental data have been collected or published on the dissolution of spent fuel in solutions saturated with dissolved hydrogen. In the SFS project there are also several planned experiments of this type with different solids (alpha-doped UO 2 , high burnup spent fuel or MOX) or solution compositions (distilled water, low ionic strength carbonated solutions, concentrated NaCl solutions). There have been already also different hypothesis forwarded to explain the data as well as full models proposed including the influence of the dissolved Fe(II) on the fuel dissolution. Some ideas towards the main lines of modeling spent fuel dissolution under such conditions will be presented. The hydrogen effect on spent fuel dissolution is relatively recent and experiments are still carried out to confirm or rule it out for different spent fuels and conditions. For this reason it would be too ambitious at the present level of knowledge to present a full modeling of such data. This is because a spent fuel dissolution model should be valid for predictions of geological time scales based on relatively short time experiments. This is possible only with a very good understanding of the dissolution process and of the mechanisms underlying the hydrogen effect, while a simple extrapolation of experimental data for repository time scales would not be reliable. (Author)

  4. Remote repair robots for dissolvers in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Sugiyama, Sen; Hirose, Yasuo; Kawamura, Hironobu; Minato, Akira; Ozaki, Norihiko.

    1984-01-01

    In nuclear facilities, for the purpose of the reduction of radiation exposure of workers, the shortening of working time and the improvement of capacity ratio of the facilities, the technical development of various devices for remote maintenance and inspection has been advanced so far. This time, an occasion came to inspect and repair the pinhole defects occurred in spent fuel dissolving tanks in the reprocessing plant of Tokai Establishment, Power Reactor and Nuclear Fuel Development Corp. However, since the radiation environmental condition and the restricting condition due to the object of repair were extremely severe, it was impossible to cope with them using conventional robot techniques. Consequently, a repair robot withstanding high level radiation has been developed anew, which can work by totally remote operation in the space of about 270 mm inside diameter and about 6 m length. The repair robot comprises a periscope reflecting mirror system, a combined underwater and atmospheric use television, a grinder, a welder, a liquid penetrant tester and an ultrasonic flaw detector. The key points of the development were the parts withstanding high level radiation and the selection of materials, to make the mechanism small size and the realization of totally remote operation. (Kako, I.)

  5. Method for separation of Cs from acid solution dissolving radionuclides and microanalysis of solution with ICP-AES

    International Nuclear Information System (INIS)

    Kanazawa, Toru; Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Fuketa, Toyoshi

    2004-06-01

    The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program is being performed at JAERI to understand mechanisms of radionuclides release from irradiated fuel during severe accidents. As a part of evaluation in the program, the mass balances of released and deposited FP (Fission Products) onto the test apparatus are estimated from gamma ray measurement for acid solution leached from the apparatus, but short-life nuclides are difficult to be measured because those in the VEGA fuel have been mostly depleted due to cooling for several years. Moreover, the radionuclides without emitting gamma rays and very small quantity of elements cannot be quantified by gamma ray measurement. Therefore, a microanalysis by ICP-AES (Inductively Coupled Plasma - Atomic Emission Spectrometry) for the acid solution leached from VEGA apparatuses is being applied to evaluate the released and deposited masses for those elements. Since Cs-134 and -137, which are major FP dissolved in the solution, have high intensity of gamma ray spectrum, they have to be removed from the solution before the microanalysis in order to avoid contamination of ICP system and to decrease exposure to gamma ray. In this report, methods for separation of Cs from acid solution were reviewed and the applicability of them to the ICP-AES analysis was discussed. The method for Cs separation using the inorganic ion exchanger, AMP (Ammonium Molybdate Phosphate) was applied to the solutions of cold and hot test and the effectiveness was examined. The results showed that more than 99.9% of Cs could be removed from the test solutions, and once removed Sb by AMP was recovered by using a complexing agent such as citric acid. Next, the method was applied to an acid solution leached from VEGA-3 apparatus, and ICP-AES analysis was performed for it. The analysis showed that amount of U, Sr and Zr were successfully quantified. Most of elements to be analyzed were measurable except for Sb, Ag and Sn, although

  6. Areva solutions for management of defective fuel

    International Nuclear Information System (INIS)

    Morlaes, I.; Vo Van, V.

    2014-01-01

    Defective fuel management is a major challenge for nuclear operators when all fuel must be long-term managed. This paper describes AREVA solutions for managing defective fuel. Transport AREVA performs shipments of defective fuel in Europe and proposes casks that are licensed for that purpose in Europe and in the USA. The paper presents the transport experience and the new European licensing approach of defective fuel transport. Dry Interim Storage AREVA is implementing the defective fuel storage in the USA, compliant with the Safety Authority's requirements. In Europe, AREVA is developing a new, more long-term oriented storage solution for defective fuel, the best available technology regarding safety requirements. The paper describes these storage solutions. Treatment Various types of defective fuel coming from around the world have been treated in the AREVA La Hague plant. Specific treatment procedures were developed when needed. The paper presents operational elements related to this experience. (authors)

  7. Impact of solute concentration on the electrocatalytic conversion of dissolved gases in buffered solutions

    KAUST Repository

    Shinagawa, Tatsuya

    2015-04-24

    To maintain local pH levels near the electrode during electrochemical reactions, the use of buffer solutions is effective. Nevertheless, the critical effects of the buffer concentration on electrocatalytic performances have not been discussed in detail. In this study, two fundamental electrochemical reactions, oxygen reduction reaction (ORR) and hydrogen oxidation reaction (HOR), on a platinum rotating disk electrode are chosen as model gas-related aqueous electrochemical reactions at various phosphate concentrations. Our detailed investigations revealed that the kinetic and limiting diffusion current densities for both the ORR and HOR logarithmically decrease with increasing solute concentration (log|jORR|=-0.39c+0.92,log|jHOR|=-0.35c+0.73). To clarify the physical aspects of this phenomenon, the electrolyte characteristics are addressed: with increasing phosphate concentration, the gas solubility decrease, the kinematic viscosity of the solution increase and the diffusion coefficient of the dissolved gases decrease. The simulated limiting diffusion currents using the aforementioned parameters match the measured ones very well (log|jORR|=-0.43c+0.99,log|jHOR|=-0.40c+0.54), accurately describing the consequences of the electrolyte concentration. These alterations of the electrolyte properties associated with the solute concentration are universally applicable to other aqueous gas-related electrochemical reactions because the currents are purely determined by mass transfer of the dissolved gases. © 2015 The Authors.

  8. Safe and reliable fuel solutions

    International Nuclear Information System (INIS)

    2013-01-01

    Published by AREVA, this booklet highlights the main aspects regarding fuel-related activities within this company. It outlines the efforts to improve all the involved processes, briefly describes the components and structure of fuel assemblies, gives an overview of Areva's different activities related to nuclear fuels (design, variety of products, fabrication, services). It outlines the relationship with the client for each of these activities, briefly describes the different parts of a fuel assembly for a PWR, outlines the importance given to quality for the fabrication processes, and indicates the different services provided by AREVA to its clients (handling, maintenance, controls, inspection, repair, training, etc.)

  9. Nondestructive determination of residual fuel on leached hulls and dissolver sludges from LWR fuel reprocessing

    International Nuclear Information System (INIS)

    Wuerz, H.; Wagner, K.; Becker, H.J.

    1990-01-01

    In reprocessing plants leached hulls and dissolver sludges represent rather important intermediate level α-waste streams. A control of the Pu content of these waste streams is desirable. The nondestructive assay method to be preferred would be passive neutron counting. However, before any decision on passive neutron monitoring becomes possible a characterization of hulls and sludges in terms of Pu content and neutron emission is necessary. For the direct determination of plutonium on hulls and in sludges, as coming from reprocessing, an active neutron measurement is required. A simple, and sufficiently sensitive active neutron method which can easily be installed uses as stationary Cf-252 neutron source. This method was used for the characterization of hulls and sludges in terms of plutonium content and total neutron emission in the WAK. Meanwhile a total of 28 batches of leached hulls and 22 batches of dissolver sludges from reprocessing of PWR fuel have been assayed. The paper describes the assay method used and gives an analysis of the error sources together with a discussion of the results and the accuracies obtained in a reprocessing plant. (orig./HP)

  10. Burner and dissolver off-gas treatment in HTR fuel reprocessing

    International Nuclear Information System (INIS)

    Barnert-Wiemer, H.; Heidendael, M.; Kirchner, H.; Merz, E.; Schroeder, G.; Vygen, H.

    1979-01-01

    In the reprocessing of HTR fuel, essentially all of the gaseous fission products are released during the heat-end tratment, which includes burning of the graphite matrix and dissolving of the heavy metallic residues in THOREX reagent. Three facilities for off-gas cleaning are described, the status of the facility development and test results are reported. Hot tests with a continuous dissolver for HTR-type fuel (throughput 2 kg HM/d) with a closed helium purge loop have been carried out. Preliminary results of these experiments are reported

  11. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  12. Trends in soil solution dissolved organic carbon (DOC) concentrations across European forests

    NARCIS (Netherlands)

    Camino-Serrano, Marta; Graf Pannatier, Elisabeth; Vicca, Sara; Luyssaert, Sebastiaan; Jonard, Mathieu; Ciais, Philippe; Guenet, Bertrand; Gielen, Bert; Peñuelas, Josep; Sardans, Jordi; Waldner, Peter; Sawicka, Kasia

    2016-01-01

    Dissolved organic carbon (DOC) in surface waters is connected to DOC in soil solution through hydrological pathways. Therefore, it is expected that long-term dynamics of DOC in surface waters reflect DOC trends in soil solution. However, a multitude of site studies have failed so far to establish

  13. Trends in soil solution dissolved organic carbon (DOC) concentrations across European forests

    NARCIS (Netherlands)

    Camino-Serrano, M.; Graf Pannatier, E.; Vicca, S.; Luyssaert, S.; Jonard, M.; Ciais, P.; Guenet, B.; Gielen, B.; Peñuelas, J.; Sardans, J.; Waldner, P.; Etzold, S.; Cecchini, G.; Clarke, N.; Galić, Z.; Gandois, L.; Hansen, K.; Johnson, J.; Klinck, U.; Lachmanová, Z.; Lindroos, A.J.; Meesenburg, H.; Nieminen, T.M.; Sanders, T.G.M.; Sawicka, K.; Seidling, W.; Thimonier, A.; Vanguelova, E.; Verstraeten, A.; Vesterdal, L.; Janssens, I.A.

    2016-01-01

    Dissolved organic carbon (DOC) in surface waters is connected to DOC in soil solution through hydrological pathways. Therefore, it is expected that long-term dynamics of DOC in surface waters reflect DOC trends in soil solution. However, a multitude of site studies have failed so far to establish

  14. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  15. Asymptotic Solutions of Serial Radial Fuel Shuffling

    Directory of Open Access Journals (Sweden)

    Xue-Nong Chen

    2015-12-01

    Full Text Available In this paper, the mechanism of traveling wave reactors (TWRs is investigated from the mathematical physics point of view, in which a stationary fission wave is formed by radial fuel drifting. A two dimensional cylindrically symmetric core is considered and the fuel is assumed to drift radially according to a continuous fuel shuffling scheme. A one-group diffusion equation with burn-up dependent macroscopic coefficients is set up. The burn-up dependent macroscopic coefficients were assumed to be known as functions of neutron fluence. By introducing the effective multiplication factor keff, a nonlinear eigenvalue problem is formulated. The 1-D stationary cylindrical coordinate problem can be solved successively by analytical and numerical integrations for associated eigenvalues keff. Two representative 1-D examples are shown for inward and outward fuel drifting motions, respectively. The inward fuel drifting has a higher keff than the outward one. The 2-D eigenvalue problem has to be solved by a more complicated method, namely a pseudo time stepping iteration scheme. Its 2-D asymptotic solutions are obtained together with certain eigenvalues keff for several fuel inward drifting speeds. Distributions of the neutron flux, the neutron fluence, the infinity multiplication factor kinf and the normalized power are presented for two different drifting speeds.

  16. A study on the expulsion of iodine from spent-fuel solutions

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Tsutomu; Takahashi, Akira; Ishikawa, Niroh [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

    1995-02-01

    During dissolution of spent nuclear fuels, some radioiodine remains in spent-fuel solutions. Its expulsion to dissolver off-gas is important to minimize iodine escape to the environment. In our current work, the iodine remaining in spent-fuel solutions varied from 0 to 10% after dissolution of spent PWR-fuel specimens (approximately 3 g each). The amount remaining probably was dependent upon the dissolution time required. The cause is ascribable to the increased nitrous acid concentration that results from NOx generated during dissolution. The presence of nitrous acid was confirmed spectrophotometrically in an NO-HNO{sub 3} system at 100{degrees}C. Experiments examining NOx concentration versus the quantity of iodine in a simulated spent-fuel solution indicate that iodine (I{minus}) in spent fuels is subjected to the following three reactions: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid arising from NOx, and (3) formation of colloidal iodine (AgI, PdI{sub 2}), the major iodine species in a spent-fuel solution. Reaction (2) competes with reaction (3) to control the quantity of iodine remaining in solution. The following two-step expulsion process to remove iodine from a spent-fuel solution was derived from these experiments: Step One - Heat spent-fuel solutions without NOx sparging. When aged colloidal iodine is present, an excess amount of iodate should be added to the solution. Step Two - Sparge the fuel solution with NOx while heating. Effect of this new method was confirmed by use of a spent PWR-fuel solution.

  17. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K [and others

    2005-03-01

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

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

    International Nuclear Information System (INIS)

    Carbol, P.; Spahiu, K.

    2005-03-01

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

  19. Improved arterial blood oxygenation following intravenous infusion of cold supersaturated dissolved oxygen solution.

    Science.gov (United States)

    Grady, Daniel J; Gentile, Michael A; Riggs, John H; Cheifetz, Ira M

    2014-01-01

    One of the primary goals of critical care medicine is to support adequate gas exchange without iatrogenic sequelae. An emerging method of delivering supplemental oxygen is intravenously rather than via the traditional inhalation route. The objective of this study was to evaluate the gas-exchange effects of infusing cold intravenous (IV) fluids containing very high partial pressures of dissolved oxygen (>760 mm Hg) in a porcine model. Juvenile swines were anesthetized and mechanically ventilated. Each animal received an infusion of cold (13 °C) Ringer's lactate solution (30 mL/kg/hour), which had been supersaturated with dissolved oxygen gas (39.7 mg/L dissolved oxygen, 992 mm Hg, 30.5 mL/L). Arterial blood gases and physiologic measurements were repeated at 15-minute intervals during a 60-minute IV infusion of the supersaturated dissolved oxygen solution. Each animal served as its own control. Five swines (12.9 ± 0.9 kg) were studied. Following the 60-minute infusion, there were significant increases in PaO2 and SaO2 (P < 0.05) and a significant decrease in PaCO2 (P < 0.05), with a corresponding normalization in arterial blood pH. Additionally, there was a significant decrease in core body temperature (P < 0.05) when compared to the baseline preinfusion state. A cold, supersaturated dissolved oxygen solution may be intravenously administered to improve arterial blood oxygenation and ventilation parameters and induce a mild therapeutic hypothermia in a porcine model.

  20. Analysis of the international criticality benchmark no 19 of a realistic fuel dissolver

    International Nuclear Information System (INIS)

    Smith, H.J.; Santamarina, A.

    1991-01-01

    The dispersion of the order of 12000 pcm in the results of the international criticality fuel dissolver benchmark calculation, exercise OECD/19, showed the necessity of analysing the calculational methods used in this case. The APOLLO/PIC method developed to treat this type of problem permits us to propose international reference values. The problem studied here, led us to investigate two supplementary parameters in addition to the double heterogeneity of the fuel: the reactivity variation as a function of moderation and the effects of the size of the fuel pellets during dissolution. The following conclusions were obtained: The fast cross-section sets used by the international SCALE package introduces a bias of - 3000 pcm in undermoderated lattices. More generally, the fast and resonance nuclear data in criticality codes are not sufficiently reliable. Geometries with micro-pellets led to an underestimation of reactivity at the end of dissolution of 3000 pcm in certain 1988 Sn calculations; this bias was avoided in the up-dated 1990 computation because of a correct use of calculation tools. The reactivity introduced by the dissolved fuel is underestimated by 3000 pcm in contributions based on the standard NITAWL module in the SCALE code. More generally, the neutron balance analysis pointed out that standard ND self shielding formalism cannot account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The combination of these three types of bias explain the underestimation of all of the international contributions of the reactivity of dissolver lattices by -2000 to -6000 pcm. The improved 1990 calculations confirm the need to use rigorous methods in the calculation of systems which involve the fuel double heterogeneity. This study points out the importance of periodic benchmarking exercises for probing the efficacity of criticality codes, data libraries and the users

  1. Process development for fabrication of zircaloy- 4 of dissolver assembly for spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Tonpe, Sunil; Saibaba, N.; Jairaj, R.N.; Ravi Shankar, A.; Kamachi Mudali, U.; Raj, Baldev

    2010-01-01

    Spent fuel reprocessing for fast breeder reactor (FBR) requires a dissolver made of a material which has resistance to corrosion as the process involves Nitric Acid as the process medium. Various materials to achieve minimum corrosion rates have been tried for this operation. Particularly the focus was on the use of advanced materials with high performance (corrosion rate and product life) for high concentrations greater than 8 N and temperatures (boiling and vapour) of Nitric Acid employed in the dissolver unit. The different commercially available materials like SS316L , Pure Titanium, Ti - 5% Ta and Ti - 5% Ta - 1.8% Nb were tried and the corrosion behavior of these materials was studied in detail. As this is continuous process of evolution of new materials, it was decided to try out zircaloy - 4 as the material of construction for construction due to its excellent corrosion resistance properties in Nitric Acid environment. The specifications were stringent and the geometrical configurations of the assembly were very intricate in shape. On accepting the challenge of fabrication of dissolver, NFC has made different fixtures for Electron Beam Welding and TIG Welding. Various trials were carried out for optimization of various operating parameter like beam current, Acceleration voltage, welding speed to get adequate weld penetration. Both EB welding and TIG welding process were standardized and qualified by carrying out a number of trials and testing these welds by various weld qualification procedures like radiography, Liquid dye penetrant testing etc. for different intricate weld geometries. All the welds were simulated with samples to optimize the weld parameters. Tests such as include metallographic (for microstructure and HAZ), mechanical (for weld strength) and chemical (material analysis for gases) were conducted and all the weld samples met the acceptable criteria. Finally the dissolver was made meeting stringent specifications. All the welds were checked

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

    International Nuclear Information System (INIS)

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

    1984-02-01

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

  3. Development of remote repair robots for dissolvers in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Sugiyama, Sen; Kunikata, Michio; Kawamura, Hironobu.

    1985-01-01

    For nuclear facilities, various types of remote maintenance and inspection devices have been developed to reduce radiation exposure to workers, save labor, and improve the operating rate of the plant. Existing robot technology, however, could not be employed when we were recently called upon to inspect and repair pinhole defects which had occurred in the spent fuel dissolvers of the Power Reactor and Nuclear Fuel Development Corporation's Tokai Reprocessing Plant, because the work had to be done in an extremely radioactive environment, conditions too extreme for conventional robots. For this reason, we developed highly radiation-resistant repair robots capable of fully remote-controlled operation inside the barrels of the dissolvers, which have the inconvenient shape of 270 mm inside diameter and 6 m length. The process for developing the six different repair robots and the their functions are described in this paper. This development was sponsored by the Power Reactor and Nuclear Fuel Development Corporation (PNC) under contract with Hitachi, Ltd. (author)

  4. [Dissolved aluminum and organic carbon in soil solution under six tree stands in Lushan forest ecosystems].

    Science.gov (United States)

    Wang, Lianfeng; Pan, Genxing; Shi, Shengli; Zhang, Lehua; Huang, Mingxing

    2003-10-01

    Different depths of soils under 6 tree stands in Lushan Botany Garden were sampled and water-digested at room temperature. The dissolved aluminum and organic carbon were then determined by colorimetry, using 8-hydroxylquilin and TOC Analyzer, respectively. The results indicated that even derived from a naturally identical soil type, the test soils exhibited a diverse solution chemistry, regarding with the Al speciation. The soil solutions under Japanese cedar, giant arborvitae and tea had lower pH values and higher contents of soluble aluminum than those under Giant dogwood, azalea and bamboo. Under giant arborvitae, the lowest pH and the highest content of total soluble aluminum and monomeric aluminum were found in soil solution. There was a significant correlation between soluble aluminum and DOC, which tended to depress the accumulation of toxic monomeric aluminum. The 6 tree stands could be grouped into 2 categories of solution chemistry, according to aluminum mobilization.

  5. Wavelength Dispersive X-ray Fluorescence Analysis of Actinides in Dissolved Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, David [Parallax Research, Inc., Tallahassee, FL (United States)

    2015-10-15

    There is an urgent need for an instrument that can quickly measure the concentration of Plutonium and other Actinides mixed with Uranium in liquids containing dissolved spent fuel rods. Parallax Research, Inc. proposes to develop an x-ray spectrometer capable of measuring U, Np and Pu in dissolved nuclear fuel rod material to less than 10 ppm levels to aid in material process control for these nuclear materials. Due to system noise produced by high radioactivity, previous x-ray spectrometers were not capable of low level measurements but the system Parallax proposed has no direct path for undesired radiation to get to the detector and the detector in the proposed device is well shielded from scatter and has very low dark current. In addition, the proposed spectrometer could measure these three elements simultaneously, also measuring background positions with an energy resolution of roughly 100 eV making it possible to see a small amount of Pu that would be hidden under the tail of the U peak in energy dispersive spectrometers. Another nearly identical spectrometer could be used to target Am and Cm if necessary. The proposed spectrometer needs only a tiny sample of roughly 1 micro-liter (1 mm3) and the measurement can be done with the liquid flowing in a radiation and chemical immune quartz capillary protected by a stainless steel rod making it possible to continuously monitor the liquid or to use a capillary manifold to measure other liquid streams. Unlike other methods such as mass spectroscopy where the sample must be taken to a remote facility and might take days for turn-around, the proposed measurement should take less than an hour. This spectrometer could enable near real-time measurement of U, Pu and Np in dilute dissolved spent nuclear fuel rod streams.

  6. Effects of Dissolving Solutions on the Accuracy of an Electronic Apex Locator-Integrated Endodontic Handpiece

    Directory of Open Access Journals (Sweden)

    Yakup Ustun

    2013-01-01

    Full Text Available The effects of three dissolving agents on the accuracy of an electronic apex locator- (EAL- integrated endodontic handpiece during retreatment procedures were evaluated. The true lengths (TLs of 56 extracted incisor teeth were determined visually. Twenty teeth were filled with gutta-percha and a resin-based sealer (group A, 20 with gutta-percha and a zinc oxide/eugenol-based sealer (group B, and 16 roots were used as the control group (group C. All roots were prepared to TL. Guttasolv, Resosolv, and Endosolv E were used as the dissolving solutions. Two evaluations of the handpiece were performed: the apical accuracy during the auto reverse function (ARL and the apex locator function (EL alone. The ARL function of the handpiece gave acceptable results. There were significant differences between the EL mode measurements and the TL (P<0.05. In these comparisons, Tri Auto ZX EL mode measurements were significantly shorter than those of the TL.

  7. Vanadium in fuel oil - a new solution

    Energy Technology Data Exchange (ETDEWEB)

    Czech, N. [Siemens, Muelheim (Germany); Finckh, H. [Siemens, Erlangen (Germany)

    1998-11-01

    Hot corrosion of the hot-gas-path components due to vanadium contamination is one of the hazards associated with heavy residual oil combustion in heavy-duty gas turbines. This economically attractive oil combustion process has benefited from the recently developed vanadium inhibition technique, which is currently being tested at the Valladolid 220 MWe combined cycle plant in Mexico. The method uses atomization of a dilute aqueous solution of Epsom salt (MgSO{sub 7},7H{sub 2}O) into very small droplets which are then injected onto the flame where intensive mixing takes place. The successful use of this new technique promises extended operating periods between cleanup operations, and cost reductions from the use of inexpensive materials, as well as the ability to operate advanced gas turbines on difficult fuels, not previously feasible. (UK)

  8. Role of dissolved oxygen on the degradation mechanism of Reactive Green 19 and electricity generation in photocatalytic fuel cell.

    Science.gov (United States)

    Lee, Sin-Li; Ho, Li-Ngee; Ong, Soon-An; Wong, Yee-Shian; Voon, Chun-Hong; Khalik, Wan Fadhilah; Yusoff, Nik Athirah; Nordin, Noradiba

    2018-03-01

    In this study, a membraneless photocatalytic fuel cell with zinc oxide loaded carbon photoanode and platinum loaded carbon cathode was constructed to investigate the impact of dissolved oxygen on the mechanism of dye degradation and electricity generation of photocatalytic fuel cell. The photocatalytic fuel cell with high and low aeration rate, no aeration and nitrogen purged were investigated, respectively. The degradation rate of diazo dye Reactive Green 19 and the electricity generation was enhanced in photocatalytic fuel cell with higher dissolved oxygen concentration. However, the photocatalytic fuel cell was still able to perform 37% of decolorization in a slow rate (k = 0.033 h -1 ) under extremely low dissolved oxygen concentration (approximately 0.2 mg L -1 ) when nitrogen gas was introduced into the fuel cell throughout the 8 h. However, the change of the UV-Vis spectrum indicates that the intermediates of the dye could not be mineralized under insufficient dissolved oxygen level. In the aspect of electricity generation, the maximum short circuit current (0.0041 mA cm -2 ) and power density (0.00028 mW cm -2 ) of the air purged photocatalytic fuel cell was obviously higher than that with nitrogen purging (0.0015 mA cm -2 and 0.00008 mW cm -2 ). Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Method of dissolving metal ruthenium

    International Nuclear Information System (INIS)

    Tsuno, Masao; Soda, Yasuhiko; Kuroda, Sadaomi; Koga, Tadaaki.

    1988-01-01

    Purpose: To dissolve and clean metal ruthenium deposited to the inner surface of a dissolving vessel for spent fuel rods. Method: Metal ruthenium is dissolved in a solution of an alkali metal hydroxide to which potassium permanganate is added. As the alkali metal hydroxide used herein there can be mentioned potassium hydroxide, sodium hydroxide and lithium hydroxide can be mentioned, which is used as an aqueous solution from 5 to 20 % concentration in view of the solubility of metal ruthenium and economical merit. Further, potassium permanganate is used by adding to the solution of alkali metal hydroxide at a concentration of 1 to 5 %. (Yoshihara, H.)

  10. Nondestructive determination of residual fuel on leached hulls and dissolver sludges from LWR fuel reprocessing

    International Nuclear Information System (INIS)

    Wuerz, H.; Wagner, K.; Becker, H.J.

    1990-01-01

    In reprocessing plants leached hulls and dissolver sludges represent rather important intermediate level α-waste streams. A control of the Pu content of these waste streams is desirable. The nondestructive assay method to be preferred would be passive neutron counting. However, before any decision on passive neutron monitoring becomes possible, a characterization of hulls and sludges in terms of Pu content and neutron emission is necessary. For the direct determination of Plutonium on hulls and in sludges, as coming from reprocessing, an active neutron measurement is required. A simple, and sufficiently sensitive active neutron method which can easily be installed uses a stationary 252 Cf neutron source. This method was used for the characterization of hulls and sludges in terms of Plutonium content and total neutron emission in the Karlsruhe reprocessing plant WAK

  11. Hydrogen Fuel Cells: Part of the Solution

    Science.gov (United States)

    Busby, Joe R.; Altork, Linh Nguyen

    2010-01-01

    With the decreasing availability of oil and the perpetual dependence on foreign-controlled resources, many people around the world are beginning to insist on alternative fuel sources. Hydrogen fuel cell technology is one answer to this demand. Although modern fuel cell technology has existed for over a century, the technology is only now becoming…

  12. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

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

  13. The effect of probe choice and solution conditions on the apparent photoreactivity of dissolved organic matter.

    Science.gov (United States)

    Maizel, Andrew C; Remucal, Christina K

    2017-08-16

    Excited triplet states of dissolved organic matter ( 3 DOM) are quantified directly with the species-specific probes trans,trans-hexadienoic acid (HDA) and 2,4,6-trimethylphenol (TMP), and indirectly with the singlet oxygen ( 1 O 2 ) probe furfuryl alcohol (FFA). Although previous work suggests that these probe compounds may be sensitive to solution conditions, including dissolved organic carbon concentration ([DOC]) and pH, and may quantify different 3 DOM subpopulations, the probes have not been systematically compared. Therefore, we quantify the apparent photoreactivity of diverse environmental waters using HDA, TMP, and FFA. By conducting experiments under ambient [DOC] and pH, with standardized [DOC] and pH, and with solid phase extraction isolates, we demonstrate that much of the apparent dissimilarity in photochemical measurements is attributable to solution conditions, rather than intrinsic differences in 3 DOM production. In general, apparent quantum yields (Φ 1 O 2 ≥ Φ 3 DOM,TMP ≫ Φ 3 DOM,HDA ) and pseudo-steady state concentrations ([ 1 O 2 ] ss > [ 3 DOM] ss,TMP > [ 3 DOM] ss,HDA ) show consistent relationships in all waters under standardized conditions. However, intrinsic differences in 3 DOM photoreactivity are apparent between DOM from diverse sources, as seen in the higher Φ 1 O 2 and lower Φ 3 DOM,TMP of wastewater effluents compared with oligotrophic lakes. Additionally, while conflicting trends in photoreactivity are observed under ambient conditions, all probes observe quantum yields increasing from surface wetlands to terrestrially influenced waters to oligotrophic lakes under standardized conditions. This work elucidates how probe selection and solution conditions influence the apparent photoreactivity of environmental waters and confirms that 3 DOM or 1 O 2 probes cannot be used interchangeably in waters that vary in [DOC], pH, or DOM source.

  14. Methods for conditioning wastes from spent fuel cans and dissolver residues

    International Nuclear Information System (INIS)

    De Regge, P.; Loida, A.; Schmidt-Hansberg, T.; Sombret, C.

    1985-04-01

    Several methods for conditioning spent fuel decladding hulls or dissolver residues have been considered in various countries of the European Community. Five of these methods use embedding technique with or without prior compaction: they are based on incorporation in metallic alloys, glass, ceramics, cements and metals or graphite compounds. A sixth one consists in melting the decladding materials. The corresponding research programs have been pursued to varying states of progress with regard to demonstrating their feasibility on an industrial scale and the use of genuine wastes in bench scale experiments. The properties of the conditioned wastes have been investigated. Special attention has been paid to the corrosion resistance to various aqueous media as tap water, brine or clayey water. Although no categorical conclusion can be drawn from the initial results, the available finding provide a basis for assessing the different processes [fr

  15. Evaluation of Purging Solutions for Military Fuel Tanks

    National Research Council Canada - National Science Library

    Rhee, In-Sik

    2003-01-01

    .... It is also a biodegradable water based solvent. Because of this property, US Army has used this environmentally friendly solvent as a purging solution in all military fuel tanks including Heavy Expanded Mobility Truck (HEMTT...

  16. Coconut endocarp and mesocarp as both biosorbents of dissolved hydrocarbons in fuel spills and as a power source when exhausted.

    Science.gov (United States)

    Luis-Zarate, Victor Hugo; Rodriguez-Hernandez, Mayra Cecilia; Alatriste-Mondragon, Felipe; Chazaro-Ruiz, Luis Felipe; Rangel-Mendez, Jose Rene

    2018-04-01

    Health and environmental problems associated with the presence of toxic aromatic compounds in water from oil spills have motivated research to develop effective and economically viable strategies to remove these pollutants. In this work, coconut shell (endocarp), coconut fiber (mesocarp) and coconut shell with fiber (endocarp and mesocarp) obtained from coconut (Cocos nucifera) waste were evaluated as biosorbents of benzene, toluene and naphthalene from water, considering the effect of the solution pH (6-9) and the presence of dissolved organic matter (DOM) in natural water (14 mg/L). In addition, the heat capacity of saturated biosorbents was determined to evaluate their potential as an alternative power source to conventional fossil fuels. Tests of N 2 physisorption, SEM, elemental and fiber analysis, ATR-FTIR and acid-based titrations were performed in order to understand the materials' characteristics, and to elucidate the biosorbents' hydrocarbon adsorption mechanism. Coconut fiber showed the highest adsorption capacities (222, 96 and 5.85 mg/g for benzene, toluene and naphthalene, respectively), which was attributed to its morphologic characteristics and to its high concentration of phenolic groups, associated with the lignin structure. The pH of the solution did not have a significant influence on the removal of the contaminants, and the presence of DOM improved the adsorption capacities of aromatic hydrocarbons. The adsorption studies showed biphasic isotherms, which highlighted the strong affinity between the molecules adsorbed on the biosorbents and the aromatic compounds remaining in the solution. Finally, combustion heat analysis of coconut waste saturated with soluble hydrocarbons showed that the heat capacity increased from 4407.79 cal/g to 5064.43 ± 11.6 cal/g, which is comparable with that of woody biomass (3400-4000 cal/g): this waste biomass with added value could be a promising biofuel. Copyright © 2018 Elsevier Ltd. All rights

  17. Photoproduction of hydrogen peroxide in aqueous solution from model compounds for chromophoric dissolved organic matter (CDOM)

    International Nuclear Information System (INIS)

    Clark, Catherine D.; Bruyn, Warren de; Jones, Joshua G.

    2014-01-01

    Highlights: • CDOM produces hydrogen peroxide in sunlit surface waters. • Quinone moieties have been proposed as the photo-active chromophore in CDOM. • Hydrogen peroxide is produced in irradiated aqueous quinone solutions. • Concentrations and production rates are comparable to humic and fulvic acids. • Optical properties post-irradiation were similar to CDOM. - Abstract: To explore whether quinone moieties are important in chromophoric dissolved organic matter (CDOM) photochemistry in natural waters, hydrogen peroxide (H 2 O 2 ) production and associated optical property changes were measured in aqueous solutions irradiated with a Xenon lamp for CDOM model compounds (dihydroquinone, benzoquinone, anthraquinone, napthoquinone, ubiquinone, humic acid HA, fulvic acid FA). All compounds produced H 2 O 2 with concentrations ranging from 15 to 500 μM. Production rates were higher for HA vs. FA (1.32 vs. 0.176 mM h −1 ); values ranged from 6.99 to 0.137 mM h −1 for quinones. Apparent quantum yields (Θ app ; measure of photochemical production efficiency) were higher for HA vs. FA (0.113 vs. 0.016) and ranged from 0.0018 to 0.083 for quinones. Dihydroquinone, the reduced form of benzoquinone, had a higher production rate and efficiency than its oxidized form. Post-irradiation, quinone compounds had absorption spectra similar to HA and FA and 3D-excitation–emission matrix fluorescence spectra (EEMs) with fluorescent peaks in regions associated with CDOM

  18. Enzymatic hydrolysis of cellulose dissolved in N-methyl morpholine oxide/water solutions.

    Science.gov (United States)

    Ramakrishnan, S; Collier, J; Oyetunji, R; Stutts, B; Burnett, R

    2010-07-01

    In situ hydrolysis of cellulose (dissolving pulp) in N-methyl morpholine oxide (NMMO) solutions by commercially available Accellerase1000 is carried out. The yield of reducing sugars is followed as a function of time at three different temperatures and four different enzyme loadings to study the effect of system parameters on enzymatic hydrolysis. Initial results show that rates of hydrolysis of cellulose and yields of reducing sugars in the presence of NMMO-water is superior initially (ratio of initial reaction rates approximately 4) and comparable to that of regenerated cellulose (for times greater than 5h) when suspended in aqueous solutions. The usage of Accellerase1000 results predominantly in the formation of glucose with minimal amounts of cellobiose. This study proves the ability of cellulases to remain active in NMMO to carry out an in situ saccharification of cellulose thus eliminating the need to recover regenerated cellulose. Thus this work will form the basis for developing a continuous process for conversion of biomass to hydrogen, ethanol and other hydrocarbons. Copyright 2009 Elsevier Ltd. All rights reserved.

  19. Surface modification of polystyrene with atomic oxygen radical anions-dissolved solution

    International Nuclear Information System (INIS)

    Wang Lian; Yan Lifeng; Zhao Peitao; Torimoto, Yoshifumi; Sadakata, Masayoshi; Li Quanxin

    2008-01-01

    A novel approach to surface modification of polystyrene (PS) polymer with atomic oxygen radical anions-dissolved solution (named as O - water) has been investigated. The O - water, generated by bubbling of the O - (atomic oxygen radical anion) flux into the deionized water, was characterized by UV-absorption spectroscopy and electron paramagnetic resonance (EPR) spectroscopy. The O - water treatments caused an obvious increase of the surface hydrophilicity, surface energy, surface roughness and also caused an alteration of the surface chemical composition for PS surfaces, which were indicated by the variety of contact angle and material characterization by atomic force microscope (AFM) imaging, field emission scanning electron microscopy (FESEM), X-ray photoelectron spectroscopy (XPS), and attenuated total-reflection Fourier transform infrared (ATR-FTIR) measurements. Particularly, it was found that some hydrophilic groups such as hydroxyl (OH) and carbonyl (C=O) groups were introduced onto the polystyrene surfaces via the O - water treatment, leading to the increases of surface hydrophilicity and surface energy. The active oxygen species would react with the aromatic ring molecules on the PS surfaces and decompose the aromatic compounds to produce hydrophilic hydroxyl and carbonyl compounds. In addition, the O - water is also considered as a 'clean solution' without adding any toxic chemicals and it is easy to be handled at room temperature. Present method may suit to the surface modification of polymers and other heat-sensitive materials potentially

  20. Fuels Containing Methane of Natural Gas in Solution

    Science.gov (United States)

    Sullivan, Thomas A.

    2004-01-01

    While exploring ways of producing better fuels for propulsion of a spacecraft on the Mars sample return mission, a researcher at Johnson Space Center (JSC) devised a way of blending fuel by combining methane or natural gas with a second fuel to produce a fuel that can be maintained in liquid form at ambient temperature and under moderate pressure. The use of such a blended fuel would be a departure for both spacecraft engines and terrestrial internal combustion engines. For spacecraft, it would enable reduction of weights on long flights. For the automotive industry on Earth, such a fuel could be easily distributed and could be a less expensive, more efficient, and cleaner-burning alternative to conventional fossil fuels. The concept of blending fuels is not new: for example, the production of gasoline includes the addition of liquid octane enhancers. For the future, it has been commonly suggested to substitute methane or compressed natural gas for octane-enhanced gasoline as a fuel for internal-combustion engines. Unfortunately, methane or natural gas must be stored either as a compressed gas (if kept at ambient temperature) or as a cryogenic liquid. The ranges of automobiles would be reduced from their present values because of limitations on the capacities for storage of these fuels. Moreover, technical challenges are posed by the need to develop equipment to handle these fuels and, especially, to fill tanks acceptably rapidly. The JSC alternative to provide a blended fuel that can be maintained in liquid form at moderate pressure at ambient temperature has not been previously tried. A blended automotive fuel according to this approach would be made by dissolving natural gas in gasoline. The autogenous pressure of this fuel would eliminate the need for a vehicle fuel pump, but a pressure and/or flow regulator would be needed to moderate the effects of temperature and to respond to changing engine power demands. Because the fuel would flash as it entered engine

  1. Logistics of the research reactor fuel cycle: AREVA solutions

    International Nuclear Information System (INIS)

    Ohayon, David; Halle, Laurent; Naigeon, Philippe; Falgoux, Jean-Louis; Franck Obadia, Franck; Auziere, Philippe

    2005-01-01

    The AREVA Group Companies offer comprehensive solutions for the entire fuel cycle of Research Reactors comply with IAEA standards. CERCA and Cogema Logistics have developed a full partnership in the front end cycle. In the field of uranium CERCA and Cogema Logistics have the long term experience of the shipment from Russia, USA to the CERCA plant.. Since 1960, CERCA has manufactured over 300,000 fuel plates and 15,000 fuel elements of more than 70 designs. These fuel elements have been delivered to 40 research reactors in 20 countries. For the Back-End stage, Cogema and Cogema Logistics propose customised solutions and services for international shipments. Cogema Logistics has developed a new generation of packaging to meet the various needs and requirements of the Laboratories and Research Reactors all over the world, and complex regulatory framework. Comprehensive assistance dedicated, services, technical studies, packaging and transport systems are provided by AREVA for every step of research reactor fuel cycle. (author)

  2. Research on the influence of ozone dissolved in the fuel-water emulsion on the parameters of the CI engine

    Science.gov (United States)

    Wojs, M. K.; Orliński, P.; Kamela, W.; Kruczyński, P.

    2016-09-01

    The article presents the results of empirical research on the impact of ozone dissolved in fuel-water emulsion on combustion process and concentration of toxic substances in CI engine. The effect of ozone presence in the emulsion and its influence on main engine characteristics (power, torque, fuel consumption) and selected parameters that characterize combustion process (levels of pressures and temperatures in combustion chamber, period of combustion delay, heat release rate, fuel burnt rate) is shown. The change in concentration of toxic components in exhausts gases when engine is fueled with ozonized emulsion was also identified. The empirical research and their analysis showed significant differences in the combustion process when fuel-water emulsion containing ozone was used. These differences include: increased power and efficiency of the engine that are accompanied by reduction in time of combustion delay and beneficial effects of ozone on HC, PM, CO and NOX emissions.

  3. 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

  4. 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

  5. Domestic fuel question and the charcoal solution

    Energy Technology Data Exchange (ETDEWEB)

    Krishna Rao, E G

    1981-06-01

    Domestic fuel for cooking forms one of the basic needs of human society. In India, the pressure of this need has exceeded the regeneration potential of the growing forests which supply a large proportion of this basic need. The pressure can be greatly relieved by converting wood to charcoal before it reaches the consumer. The present paper examines this aspect and reviews the modern methods of charcoal production on fuelwood resources. Besides being a choice domestic fuel, charcoal is a valuable raw material in various industries. Charcoal making industry can be established as a rural based industry (as part of community forestry projects) and would generate much needed cash income at grassroot level. The strategy would be important in dealing with the problem of chronic poverty at this level. (Refs. 5).

  6. Photoproduction of hydrogen peroxide in aqueous solution from model compounds for chromophoric dissolved organic matter (CDOM).

    Science.gov (United States)

    Clark, Catherine D; de Bruyn, Warren; Jones, Joshua G

    2014-02-15

    To explore whether quinone moieties are important in chromophoric dissolved organic matter (CDOM) photochemistry in natural waters, hydrogen peroxide (H2O2) production and associated optical property changes were measured in aqueous solutions irradiated with a Xenon lamp for CDOM model compounds (dihydroquinone, benzoquinone, anthraquinone, napthoquinone, ubiquinone, humic acid HA, fulvic acid FA). All compounds produced H2O2 with concentrations ranging from 15 to 500 μM. Production rates were higher for HA vs. FA (1.32 vs. 0.176 mM h(-1)); values ranged from 6.99 to 0.137 mM h(-1) for quinones. Apparent quantum yields (Θ app; measure of photochemical production efficiency) were higher for HA vs. FA (0.113 vs. 0.016) and ranged from 0.0018 to 0.083 for quinones. Dihydroquinone, the reduced form of benzoquinone, had a higher production rate and efficiency than its oxidized form. Post-irradiation, quinone compounds had absorption spectra similar to HA and FA and 3D-excitation-emission matrix fluorescence spectra (EEMs) with fluorescent peaks in regions associated with CDOM. Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. A durable and dependable solution for RTR spent fuel management

    International Nuclear Information System (INIS)

    Thomasson, J.

    1999-01-01

    RTR Operators need efficient and cost-effective services for the management of their spent fuel and this, for the full lifetime of their facility. Thanks to the integration of transport, reprocessing and conditioning services, COGEMA provides a cogent solution, with the utmost respect for safety and preservation of the environment, for the short, medium and long terms. As demonstrated in this paper, this option offers the only durable and dependable solution for the RTR spent fuel management, leading to a conditioning for the final residues directly suitable for final disposal. The main advantage of such an option is obviously the significant reduction in terms of volume and radiotoxicity of the ultimate waste when compared to direct disposal of spent fuels. The efficiency of such a solution has been proven, some RTR operators having already trusted COGEMA for the management of their aluminide fuel. With its commitment in R and D activities for the development of a high performance and reprocessable LEU fuels, COGEMA will be able to propose a solution for all types of fuels, HEU and LEU

  8. The effect of CO{sub 2} dissolved in a diesel fuel on the jet flame characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Xiao Jin; Huang Zhen; Qiao Xinqi; Hou Yuchun [Shanghai Jiao Tong University, Shanghai (China). Research Institute of Internal Combustion Engine

    2008-03-15

    This paper is concerned with an experimental study of the jet diffusion flame characteristics of fuel containing CO{sub 2}. Using diesel fuel containing dissolved CO{sub 2} gas, experiments were performed under atmospheric conditions with a diesel hole-type nozzle of 0.19 mm orifice diameter at constant injection pressure. In this study, four different CO{sub 2} mass fraction in diesel fuel such as 3.13%, 7.18%, 12.33% and 17.82% were used to study the effect of CO{sub 2} concentration on the jet flame characteristics. Jet flame characteristics were measured by direct photography, meanwhile the image colorimetry is used to assess the qualitative features of jet flame temperature. Experimental results show that the CO{sub 2} gas dilution effect and the atomization effect have a great influence on the flame structure and average temperature. When the injection pressure of diesel fuel increased from 4 MPa to 6 MPa, the low temperature flame length increased from 18.4 cm to 21.7 cm and the full temperature flame length decreased from 147.6 cm to 134.7 cm. With the increase of CO{sub 2} gas dissolved in the diesel fuel, the jet flame full length decreased for the jet atomization being improved greatly meanwhile the low temperature flame length increased for the CO{sub 2} gas dilution effect; with the increase of CO{sub 2} gas dissolved in the diesel fuel, the average temperature of flame increases firstly and then falls. Experimental results validate that higher injection pressure will improve jet atomization and then increased the flame average temperature. 27 refs., 13 figs.

  9. Fuel handling solutions to power pulse at Bruce NGS A

    International Nuclear Information System (INIS)

    Day, R.C.

    1996-01-01

    In response to the discovery of the power pulse problem in March of 1993, Bruce A has installed flow straightening shield plugs in the inner zone channels of all units to partially reduce the gap and gain an increase in reactor power to 75%. After review and evaluation of solutions to manage the gap, including creep compensators and long fuel bundles, efforts have focused on a different solution involving reordering the fuel bundles to reverse the burnup profile. This configuration is maintained by fuelling with the flow and providing better support to the highly irradiated downstream fuel bundles by changing the design of the outlet shield plug. Engineering changes to the fuel handling control system and outlet shield plug are planned to be implemented starting in June 1996, thereby eliminating the power pulse problem and restrictions on reactor operating power. (author). 2 refs., 1 tab., 2 figs

  10. Fuel handling solutions to power pulse at Bruce NGS A

    Energy Technology Data Exchange (ETDEWEB)

    Day, R C [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

    1997-12-31

    In response to the discovery of the power pulse problem in March of 1993, Bruce A has installed flow straightening shield plugs in the inner zone channels of all units to partially reduce the gap and gain an increase in reactor power to 75%. After review and evaluation of solutions to manage the gap, including creep compensators and long fuel bundles, efforts have focused on a different solution involving reordering the fuel bundles to reverse the burnup profile. This configuration is maintained by fuelling with the flow and providing better support to the highly irradiated downstream fuel bundles by changing the design of the outlet shield plug. Engineering changes to the fuel handling control system and outlet shield plug are planned to be implemented starting in June 1996, thereby eliminating the power pulse problem and restrictions on reactor operating power. (author). 2 refs., 1 tab., 2 figs.

  11. Influences of dissolved oxygen concentration on biocathodic microbial communities in microbial fuel cells.

    Science.gov (United States)

    Rago, Laura; Cristiani, Pierangela; Villa, Federica; Zecchin, Sarah; Colombo, Alessandra; Cavalca, Lucia; Schievano, Andrea

    2017-08-01

    Dissolved oxygen (DO) at cathodic interface is a critical factor influencing microbial fuel cells (MFC) performance. In this work, three MFCs were operated with cathode under different DO conditions: i) air-breathing (A-MFC); ii) water-submerged (W-MFC) and iii) assisted by photosynthetic microorganisms (P-MFC). A plateau of maximum current was reached at 1.06±0.03mA, 1.48±0.06mA and 1.66±0.04mA, increasing respectively for W-MFC, P-MFC and A-MFC. Electrochemical and microbiological tools (Illumina sequencing, confocal microscopy and biofilm cryosectioning) were used to explore anodic and cathodic biofilm in each MFC type. In all cases, biocathodes improved oxygen reduction reaction (ORR) as compared to abiotic condition and A-MFC was the best performing system. Photosynthetic cultures in the cathodic chamber supplied high DO level, up to 16mg O2 L -1 , which sustained aerobic microbial community in P-MFC biocathode. Halomonas, Pseudomonas and other microaerophilic genera reached >50% of the total OTUs. The presence of sulfur reducing bacteria (Desulfuromonas) and purple non-sulfur bacteria in A-MFC biocathode suggested that the recirculation of sulfur compounds could shuttle electrons to sustain the reduction of oxygen as final electron acceptor. The low DO concentration limited the cathode in W-MFC. A model of two different possible microbial mechanisms is proposed which can drive predominantly cathodic ORR. Copyright © 2017 Elsevier B.V. All rights reserved.

  12. Siemens fuel gasification technology - solutions and developments

    Energy Technology Data Exchange (ETDEWEB)

    Hannemann, F.; Schingnitz, M.; Schmid, C. [Siemens Fuel Gasification Technology GmbH, Freiberg (Germany)

    2007-07-01

    In 2006, Siemens Power Generation Group acquired the GSP Gasification technology, and renamed it SFGT. The presentation reviews the technology and provides an update on current projects. The future plans for the development of the technology based on extensive experience and comprehensive development work gathered over many years and proven in a number of gasification plants is covered. SFGT operates, at its Freiberg facility, a 5 MWth pilot plant which was built to test prototype designs and to determine process conditions for various feed streams. An overview is given of the results of tests completed on a wide range of carbonaceous materials including all types of solid fuels from lignite to anthracite, as well as brown coal, oil, sludge or biomass, and low-temperature coke or petcoke. The technical focus of the paper is on the unique design features such as the cooling screen and alternative refractory lining, as well as the dense flow feeding system that allows the preferable use of lignite applications.

  13. K Basin Sludge Conditioning Process Testing Partitioning of PCBs in Dissolver Solution After Neutralization/Precipitation (Caustic Adjustment)

    International Nuclear Information System (INIS)

    Schmidt, A.J.; Thornton, B.M.; Hoppe, E.W.; Mong, G.M.; Silvers, K.L.; Slate, S.O.

    1999-01-01

    The purpose of the work described in this report was to gain a better understanding of how PCB congeners present in a simulated K Basin sludge dissolver solution will partition upon neutralization and precipitation (i.e., caustic adjustment). In a previous study (Mong et al. 1998),the entire series of sludge conditioning steps (acid dissolution, filtration, and caustic adjustment) were examined during integrated testing. In the work described here, the caustic adjustment step was isolated to examine the fate of PCBs in more detail within this processing step. For this testing, solutions of dissolver simulant (containing no solids) with a known initial concentration of PCB congeners were neutralized with caustic to generate a clarified supernatant and a settled sludge phase. PCBs were quantified in each phase (including the PCBs associated with the test vessel rinsates), and material balance information was collected

  14. Study of source term evaluation from fuel solution under simulated nuclear criticality accident in TRACY

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Tashiro, Shinsuke; Nagai, Hitoshi; Koike, Tadao; Okagawa, Seigo; Murata, Mikio

    1999-01-01

    In a accident at the dissolver in a reprocessing plant, various fission products and radiolysis gases will be produced in the fuel solution and volatile radioactive nuclides and radiolysis gases and nitrogen oxide will be released into vent-gas spontaneously. Moreover other on-volatile nuclide will be releases as radioactive aerosol (mist) with bursting bubbles at surface of the solution. Therefore quantitative estimation of release and transport behavior of the radioactive material from solution as source term is very important. TRACY is a transient criticality experimental facility for studying the transient criticality characteristics of low enriched uranium. In this paper, experiment methods and results about the release behavior of the hydrogen, radioactive aerosol and iodine species from the fuel solutions are reported. As the results of the experiments, release patterns of H 2 , 140 Ba and 131 I could be grasped. Concentrations of H 2 in the vent-gas and 140 Ba in the gas phase in the core tank attained to the peak just after the transient criticality and decreased exponentially with time. On the other hand, concentrations of 131 I in the gas phase of the tank began to increase with a time lag of several minutes from the transient criticality and attained approximately constant values. (J.P.N.)

  15. Solution to a fuel-and-cladding rewetting model

    International Nuclear Information System (INIS)

    Olek, S.

    1989-06-01

    A solution by the Wiener-Hopf technique is derived for a model for the rewetting of a nuclear fuel rod. The gap between the fuel and the cladding is modelled by an imperfect contact between the two. A constant heat transfer coefficient is assumed on the wet side, whereas the dry side is assumed to be adiabatic. The solution for the rewetting temperature is in the form of an integral whose integrand contains the model parameters, including the rewetting velocity. Numerical results are presented for a large number of these parameters. It is shown that there are such large values of the rewetting temperature and the gap resistance, or such low values of the initial wall temperature, for which the rewetting velocity is unaffected by the fuel properties. (author) l fig., 7 tabs., 17 refs

  16. Radiation Re-solution Calculation in Uranium-Silicide Fuels

    International Nuclear Information System (INIS)

    Matthews, Christopher; Andersson, Anders David Ragnar; Unal, Cetin

    2017-01-01

    The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can cause collisions in the bubble that result in gas atoms being knocked back into the grain. This so called ''re-solution'' event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.

  17. Reclamation of cadmium-contaminated soil using dissolved organic matter solution originating from wine-processing waste sludge

    International Nuclear Information System (INIS)

    Liu, Cheng-Chung; Chen, Guan-Bu

    2013-01-01

    Highlights: ► Increases in acidity, washing frequency, and operational temperature enhance the Cd removal. ► Approximately 80% of Cd can be removed from the soil by dissolved organic matter (DOM) washing. ► The DOM washing can moderate the loss of soil fertility. ► The DOM washing will have a great improvement if we employ NaOH, KOH, Ca(OH) 2 , and Mg(OH) 2 to prepare the DOM solution together. -- Abstract: Soil washing using an acid solution is a common practice for removing heavy metals from contaminated soil in Taiwan. However, serious loss of nutrients from soil is a major drawback of the washing. Distillery sludge can be used to prepare a dissolved organic matter (DOM) solution by extracting its organic constituents with alkaline solutions. This study employed DOM solutions to remediate Cd-contaminated soil (with concentrations up to 21.5 mg kg −1 ) and determine the factors affecting removal of Cd, such as pH, initial concentration of DOM solution, temperature, and washing frequency. When washing with pH 3.0 and 1250 mg L −1 DOM solution, about 80% and 81% of Cd were removed from the topsoil at 27 °C and subsoil at 40 °C, respectively. To summarize the changes in fertility during DOM washing with various pH solutions: the increase in organic matter content ranged from 7.7% to 23.7%; cation exchange capacity (CEC) ranged from 4.6% to 13.9%; available ammonium (N-NH 4 ) content ranged from 39.4% to 2175%; and available phosphorus content ranged from 34.5% to 182%. Exchangeable K, Ca, and Mg remained in the topsoil after DOM washing, with concentrations of 1.1, 2.4, and 1.5 times higher than those treated with HCl solution at the same pH, respectively

  18. Supply Chain-based Solution to Prevent Fuel Tax Evasion

    Energy Technology Data Exchange (ETDEWEB)

    Franzese, Oscar [ORNL; Capps, Gary J [ORNL; Daugherty, Michael [United States Department of Transportation (USDOT), Federal Highway Administration (FHWA); Siekmann, Adam [ORNL; Lascurain, Mary Beth [ORNL; Barker, Alan M [ORNL

    2016-01-01

    The primary source of funding for the United States transportation system is derived from motor fuel and other highway use taxes. Loss of revenue attributed to fuel-tax evasion has been assessed to be somewhere between $1 billion per year, or approximately 25% of the total tax collected. Any solution that addresses this problem needs to include not only the tax-collection agencies and auditors, but also the carriers transporting oil products and the carriers customers. This paper presents a system developed by the Oak Ridge National Laboratory for the Federal Highway Administration which has the potential to reduce or eliminate many fuel-tax evasion schemes. The solution balances the needs of tax-auditors and those of the fuel-hauling companies and their customers. The technology was deployed and successfully tested during an eight-month period on a real-world fuel-hauling fleet. Day-to-day operations of the fleet were minimally affected by their interaction with this system. The results of that test are discussed in this paper.

  19. Linking variability in soil solution dissolved organic carbon to climate, soil type, and vegetation type

    NARCIS (Netherlands)

    Camino-Serrano, Marta; Gielen, Bert; Luyssaert, Sebastiaan; Ciais, Philippe; Vicca, Sara; Guenet, Bertrand; Vos, Bruno De; Cools, Nathalie; Ahrens, Bernhard; Altaf Arain, M.; Borken, Werner; Clarke, Nicholas; Clarkson, Beverley; Cummins, Thomas; Don, Axel; Pannatier, Elisabeth Graf; Laudon, Hjalmar; Moore, Tim; Nieminen, Tiina M.; Nilsson, Mats B.; Peichl, Matthias; Schwendenmann, Luitgard; Siemens, Jan; Janssens, Ivan

    2014-01-01

    Lateral transport of carbon plays an important role in linking the carbon cycles of terrestrial and aquatic ecosystems. There is, however, a lack of information on the factors controlling one of the main C sources of this lateral flux, i.e., the concentration of dissolved organic carbon (DOC) in

  20. Sustainable Solutions for Nuclear used Fuels Interim Storage

    International Nuclear Information System (INIS)

    Arslan, Marc; Favet, Dominique; Issard, Herve; Le Jemtel, Amaury; Drevon, Caroline

    2014-01-01

    AREVA has a unique experience in providing sustainable solutions for used fuel management, fitted with the needs of different customers in the world and with regulation in different countries. These solutions entail both recycling and interim storage technologies. In a first part, we will describe the various types of solutions for Interim Storage of UNF that have been implemented around the world for interim storage at reactor or centralized Pad solution in canisters dry storage, vault type storages for dry storage, dry storage of transportation casks (dual purpose) pools for wet storage, The experience for all these different families of interim storages in which AREVA is involved is extensive and will be discussed with respect to the new challenges: increase of the duration of the interim storage (long term interim storage) increase of burn up of the fuels In a second part of the presentation, special recycling features will be presented. In that case, interim storage of the used fuels is ensured in pools. This provides in the long term good conditions for the behaviour of the fuel and its retrievability. With recycling, the final waste (Universal Canister of vitrified fission products and compacted hulls and end pieces): is stable and licensed in many countries for the final disposal (France, UK, Belgium, NL, Switzerland, Germany, Japan, upcoming: Spain, Australia, Italy). Presents neither safety criticality risks nor proliferation risks (AREVA conditioned HLW and LL-ILW are free of IAEA safeguard constraints thanks to AREVA process high recovery and purification yields). It can therefore be safely stored in interim storage for more than 100 years before final disposal. Some economic considerations will also be discussed. In particular, in the case of long term interim storage of used fuels, there are growing uncertainties regarding the future needs of repackaging and transportation, which can result in future cost overruns. Meanwhile, in the recycling policy

  1. Reactivity effect of non-uniformly distributed fuel in fuel solution systems

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Yamane, Yoshihiro; Nishina, Kojiro; Mitsuhashi, Ishi.

    1991-01-01

    A numerical method to determine the optimal fuel distribution for minimum critical mass, or maximum k-effective, is developed using the Maximum Principle in order to evaluate the maximum effect of non-uniformly distributed fuel on reactivity. This algorithm maximizes the Hamiltonian directly by an iterative method under a certain constraint-the maintenance of criticality or total fuel mass. It ultimately reaches the same optimal state of a flattened fuel importance distribution as another algorithm by Dam based on perturbation theory. This method was applied to two kinds of spherical cores with water reflector in the simulating reprocessing facility. In the slightly-enriched uranyl nitrate solution core, the minimum critical mass decreased by less than 1% at the optimal moderation state. In the plutonium nitrate solution core, the k-effective increment amounted up to 4.3% Δk within the range of present study. (author)

  2. Quasi-three-dimensional analysis of ground water flow and dissolved multicomponent solute transport in saturated porous media

    International Nuclear Information System (INIS)

    Tang, Yi.

    1991-01-01

    A computational procedure was developed in this study to provide flexibility needed in the application of three-dimensional groundwater flow and dissolved multicomponent solute transport simulations. In the first part of this study, analytical solutions were proposed for the dissolved single-component solute transport problem. These closed form solutions were developed for homogeneous but stratified porous media. This analytical model took into account two-dimensional diffusion-advection in the main aquifer layer and one-dimensional diffusion-advection in the adjacent aquitards, as well as first order radioactive decay and linear adsorption isotherm in both aquifer and aquitards. The associated analytical solutions for solute concentration distributions in the aquifer and aquitards were obtained using Laplace Transformation and Method of Separation of Variables techniques. Next, in order to analyze the problem numerically, a quasi-three-dimensional finite element algorithm was developed based on the multilayer aquifer concept. In this phase, advection, dispersion, adsorption and first order multi-species chemical reaction terms were included to the analysis. Employing this model, without restriction on groundwater flow pattern in the multilayer aquifer system, one may analyze the complex behavior of the groundwater flow and solute movement pattern in the system. These numerical models may be utilized as calibration tools in site characterization studies, or as predictive models during the initial stages of a typical site investigation study. Through application to several test and field problems, the usefulness, accuracy and efficiency of the proposed models were demonstrated. Comparison of results with analytical solution, experimental data and other numerical methods were also discussed

  3. On the behaviour of dissolved fission gases prior to transient testing of fuel pins

    International Nuclear Information System (INIS)

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

    1978-10-01

    The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

  4. Accumulation of dissolved gases at hydrophobic surfaces in water and sodium chloride solutions: Implications for coal flotation

    Energy Technology Data Exchange (ETDEWEB)

    Hampton, M.A.; Nguyen, A.V. [University of Queensland, Brisbane, Qld. (Australia). Division of Chemical Engineering

    2009-08-15

    Dissolved gases can preferentially accumulate at the hydrophobic solid-water interface as revealed by neutron reflectivity measurements. In this paper, atomic force microscopy (AFM) was used to examine accumulation of dissolved gases at a hydrophobic surface in water and sodium chloride solutions. The solvent-exchange method was used to artificially form gaseous domains accumulated at the interface suitable for AFM imaging. Smooth graphite surfaces were used as model surfaces to minimize the secondary effect of surface roughness on the imaging. The concentration of NaCl up to 1 M was found to have a negligible influence on the geometry and population of pre-existing nanobubbles, nanopancakes and nanobubble-nanopancake composites. The implications of the findings on coal flotation in saline water are discussed in terms of attraction between hydrophobic surfaces in water, bubble-particle attachment and hydrophobic coagulation between particles.

  5. The extraction of trace amounts of gold from different aqueous mineral acid solutions by diphenyl-2-pyridylmethane dissolved in chloroform

    International Nuclear Information System (INIS)

    Iqbal, M.; Ejaz, M.; Chaudhri, S.A.; Zamiruddin

    1978-01-01

    Diphenyl-2-pyridylmethane, a high-molecular-weight substituted pyridine has been examined. Its behaviour is similar to that of amines in that it forms salts with mineral acids. The acid ionization constant (pKsub(BHsup(+)) is 4.4+-0.06 at 25 deg C. A study of the partition behaviour of trace amounts of gold between mineral acid solutions and 0.1 M diphenyl-2-pyridylmethane dissolved in chloroform indicates that the metal can be quantitatively extracted from dilute mineral acid solutions and also from concentrated hydrochloric acid media in a single extraction. Common anions have little effect on extraction in concentrations up to 1 M. Separation factors of a number of metal ions relative to gold are reported for three mineral acid systems. Gold has been estimated in some synthetic samples using a neutron-activation technique by prior extraction with 0.1 M solution of diphenyl-2-pyridylmethane dissolved in chloroform. Distribution of the test elements between aqueous and organic phase was followed radiometrically. The solutions (usually 1 cm 3 ) were shaken in stoppered vials for 5 minutes using a mechanical shaker. After separation of the layers, 500 μl of each phase were taken for radiochemical analysis. The standard deviation did not exceed 1%. (T.G.)

  6. Reclamation of cadmium-contaminated soil using dissolved organic matter solution originating from wine-processing waste sludge.

    Science.gov (United States)

    Liu, Cheng-Chung; Chen, Guan-Bu

    2013-01-15

    Soil washing using an acid solution is a common practice for removing heavy metals from contaminated soil in Taiwan. However, serious loss of nutrients from soil is a major drawback of the washing. Distillery sludge can be used to prepare a dissolved organic matter (DOM) solution by extracting its organic constituents with alkaline solutions. This study employed DOM solutions to remediate Cd-contaminated soil (with concentrations up to 21.5 mg kg(-1)) and determine the factors affecting removal of Cd, such as pH, initial concentration of DOM solution, temperature, and washing frequency. When washing with pH 3.0 and 1250 mg L(-1) DOM solution, about 80% and 81% of Cd were removed from the topsoil at 27 °C and subsoil at 40 °C, respectively. To summarize the changes in fertility during DOM washing with various pH solutions: the increase in organic matter content ranged from 7.7% to 23.7%; cation exchange capacity (CEC) ranged from 4.6% to 13.9%; available ammonium (NNH(4)) content ranged from 39.4% to 2175%; and available phosphorus content ranged from 34.5% to 182%. Exchangeable K, Ca, and Mg remained in the topsoil after DOM washing, with concentrations of 1.1, 2.4, and 1.5 times higher than those treated with HCl solution at the same pH, respectively. Copyright © 2012 Elsevier B.V. All rights reserved.

  7. Top-MOX fuel solution: strategies, challenges, opportunities

    International Nuclear Information System (INIS)

    Breitenstein, P.; Vo Van, V.

    2014-01-01

    TOP-MOX is a nuclear fuel solution and product developed by AREVA and successfully implemented in Europe. It allows utilities burning plutonium (instead of enriched uranium) even when this plutonium is not stemming from own reprocessed used fuel - that is third party plutonium. The important challenges for utilities along with TOP-MOX implementation are legal/patrimonial Pu-ownership issues and general economical aspects. Available sponsorship of such plutonium permits UO2 competitive market prices. For new MOX customers licensing and technical aspects come along. Further AREVA proposes a flexible solution which is called 'TOP-MOX pre-cycling'. This involves making available third party plutonium for fuel fabrication and reactor use pending the utilities' final strategic fuel cycle decision. The paper gives insight into and analyses the impacts of allowing customers the implementation of a TOP-MOX program with focus on Pu-ownership, economics, technical and legal aspects as well as the impact on used MOX management and final waste management. (authors)

  8. Ecological consequences of elevated total dissolved solids associated with fossil fuel extraction in the United States

    Science.gov (United States)

    Fossil fuel burning is considered a major contributor to global climate change. The outlook for production and consumption of fossil fuels int he US indicates continued growth to support growing energy demands. For example, coal-generated electricity is projected ot increase from...

  9. Benchmark solution of contemporary PWR integral fuel burnable absorbers

    International Nuclear Information System (INIS)

    Stucker, D.L.; Hone, M.J.; Holland, R.A.

    1993-01-01

    This paper presents a closely controlled benchmark solution of the two major contemporary pressurized water reactor integral burnable absorber designs: zirconium diboride (ZrB 2 ) and gadolinia (Gd 2 O 3 ). The comparison is accomplished using self-generating equilibrium cycles with equal energy, equal discharge burnup, and equal safety constraints. The reference plant for this evaluation is a 3411-MW(thermal) Westinghouse four-loop nuclear steam supply system operating with an inlet temperature of 285.9 degrees C, a core coolant mass now rate of 16877.3 kg/s, and coolant pressure of 15.5 MPa. The reactor consists of 193 VANTAGE 5H fuel assemblies that are discharged at a region average burnup of 48.4 GWd/tonne U. Each fuel assembly contains a natural uranium axial blanket 15.24 cm long at the top and the bottom of the fuel rod. The burnable absorber rods are symmetrically radially dispersed within the fuel assembly such that intrabundle power peaking is minimized. The burnable absorber material for both ZrB 2 and Gd 2 O 3 is axially zoned to the central 304.8 cm of the absorber-bearing fuel rods. The fuel management was constrained such that the thermal and safety limitations of F δH q -5 /degrees C were simultaneously achieved. The maximum long-term operating soluble boron concentration was also limited to 446 effective full-power days (EFPDs) including 14 EFPDs of power coastdown were assumed

  10. Reclamation of cadmium-contaminated soil using dissolved organic matter solution originating from wine-processing waste sludge

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Cheng-Chung, E-mail: ccliu@niu.edu.tw [Department of Environmental Engineering, National Ilan University, Ilan, 260, Taiwan (China); Chen, Guan-Bu [Department of Environmental Engineering, National Ilan University, Ilan, 260, Taiwan (China)

    2013-01-15

    Highlights: ► Increases in acidity, washing frequency, and operational temperature enhance the Cd removal. ► Approximately 80% of Cd can be removed from the soil by dissolved organic matter (DOM) washing. ► The DOM washing can moderate the loss of soil fertility. ► The DOM washing will have a great improvement if we employ NaOH, KOH, Ca(OH){sub 2}, and Mg(OH){sub 2} to prepare the DOM solution together. -- Abstract: Soil washing using an acid solution is a common practice for removing heavy metals from contaminated soil in Taiwan. However, serious loss of nutrients from soil is a major drawback of the washing. Distillery sludge can be used to prepare a dissolved organic matter (DOM) solution by extracting its organic constituents with alkaline solutions. This study employed DOM solutions to remediate Cd-contaminated soil (with concentrations up to 21.5 mg kg{sup −1}) and determine the factors affecting removal of Cd, such as pH, initial concentration of DOM solution, temperature, and washing frequency. When washing with pH 3.0 and 1250 mg L{sup −1} DOM solution, about 80% and 81% of Cd were removed from the topsoil at 27 °C and subsoil at 40 °C, respectively. To summarize the changes in fertility during DOM washing with various pH solutions: the increase in organic matter content ranged from 7.7% to 23.7%; cation exchange capacity (CEC) ranged from 4.6% to 13.9%; available ammonium (N-NH{sub 4}) content ranged from 39.4% to 2175%; and available phosphorus content ranged from 34.5% to 182%. Exchangeable K, Ca, and Mg remained in the topsoil after DOM washing, with concentrations of 1.1, 2.4, and 1.5 times higher than those treated with HCl solution at the same pH, respectively.

  11. Development of a continuous process for adjusting nitrate, zirconium, and free hydrofluoric acid concentrations in zirconium fuel dissolver product

    International Nuclear Information System (INIS)

    Cresap, D.A.; Halverson, D.S.

    1993-04-01

    In the Fluorinel Dissolution Process (FDP) upgrade, excess hydrofluoric acid in the dissolver product must be complexed with aluminum nitrate (ANN) to eliminate corrosion concerns, adjusted with nitrate to facilitate extraction, and diluted with water to ensure solution stability. This is currently accomplished via batch processing in large vessels. However, to accommodate increases in projected throughput and reduce water production in a cost-effective manner, a semi-continuous system (In-line Complexing (ILC)) has been developed. The major conclusions drawn from tests demonstrating the feasibility of this concept are given in this report

  12. Influence of dissolved organic matter and manganese oxides on metal speciation in soil solution: A modelling approach.

    Science.gov (United States)

    Schneider, Arnaud R; Ponthieu, Marie; Cancès, Benjamin; Conreux, Alexandra; Morvan, Xavier; Gommeaux, Maxime; Marin, Béatrice; Benedetti, Marc F

    2016-06-01

    Trace element (TE) speciation modelling in soil solution is controlled by the assumptions made about the soil solution composition. To evaluate this influence, different assumptions using Visual MINTEQ were tested and compared to measurements of free TE concentrations. The soil column Donnan membrane technique (SC-DMT) was used to estimate the free TE (Cd, Cu, Ni, Pb and Zn) concentrations in six acidic soil solutions. A batch technique using DAX-8 resin was used to fractionate the dissolved organic matter (DOM) into four fractions: humic acids (HA), fulvic acids (FA), hydrophilic acids (Hy) and hydrophobic neutral organic matter (HON). To model TE speciation, particular attention was focused on the hydrous manganese oxides (HMO) and the Hy fraction, ligands not considered in most of the TE speciation modelling studies in soil solution. In this work, the model predictions of free ion activities agree with the experimental results. The knowledge of the FA fraction seems to be very useful, especially in the case of high DOM content, for more accurately representing experimental data. Finally, the role of the manganese oxides and of the Hy fraction on TE speciation was identified and, depending on the physicochemical conditions of the soil solution, should be considered in future studies. Copyright © 2016 Elsevier Ltd. All rights reserved.

  13. Plutonium estimation in the process solutions and oxide dissolved audit samplers by potentiometry using memo titrator

    International Nuclear Information System (INIS)

    Kumaraguru, K.; Shukla, Y.D.; Vijayan, K.; Ramamoorthy, N.; Jambunathan, U.; Kapoor, S.C.

    1990-01-01

    Potentiometric method is employed by using memotitrator coupled with combined electrode for the estimation of plutonium. The estimations are carried out on the process samples and the acid dissolved samples for auditing, in the concentration range of 5 g/l to 20 g/l. The chemical procedure is: i)oxidising plutonium to higher oxidation state by silver oxide, ii)reducing the same by adding excess ferrous, and iii)titrating potassium dichromate against the unreacted ferrous. The plutonium content is computed from ferrous consumed in the reaction. The average percentage error of the method is +/-0.27. The values obtained are in close agreement with those obtained by coulometry. (author)

  14. Effect of TCE concentration and dissolved groundwater solutes on NZVI-promoted TCE dechlorination and H2 evolution.

    Science.gov (United States)

    Liu, Yueqiang; Phenrat, Tanapon; Lowry, Gregory V

    2007-11-15

    Nanoscale zero-valent iron (NZVI) is used to remediate contaminated groundwater plumes and contaminant source zones. The target contaminant concentration and groundwater solutes (NO3-, Cl-, HCO3-, SO4(2-), and HPO4(2-)) should affect the NZVI longevity and reactivity with target contaminants, but these effects are not well understood. This study evaluates the effect of trichloroethylene (TCE) concentration and common dissolved groundwater solutes on the rates of NZVI-promoted TCE dechlorination and H2 evolution in batch reactors. Both model systems and real groundwater are evaluated. The TCE reaction rate constant was unaffected by TCE concentration for [TCE] TCE concentration up to water saturation (8.4 mM). For [TCE] > or = 0.46 mM, acetylene formation increased, and the total amount of H2 evolved at the end of the particle reactive lifetime decreased with increasing [TCE], indicating a higher Fe0 utilization efficiency for TCE dechlorination. Common groundwater anions (5mN) had a minor effect on H2 evolution but inhibited TCE reduction up to 7-fold in increasing order of Cl- TCE reduction but increased acetylene production and decreased H2 evolution. NO3- present at > 3 mM slowed TCE dechlorination due to surface passivation. NO3- present at 5 mM stopped TCE dechlorination and H2 evolution after 3 days. Dissolved solutes accounted for the observed decrease of NZVI reactivity for TCE dechlorination in natural groundwater when the total organic content was small (< 1 mg/L).

  15. Effect of Dissolved Oxygen and Immersion Time on the Corrosion Behaviour of Mild Steel in Bicarbonate/Chloride Solution

    Directory of Open Access Journals (Sweden)

    Gaius Debi Eyu

    2016-09-01

    Full Text Available The electrochemical behavior of mild steel in bicarbonate solution at different dissolved oxygen (DO concentrations and immersion times has been studied under dynamic conditions using electrochemical techniques. The results show that both DO and immersion times influence the morphology of the corrosion products. In comparative tests, the corrosion rate was systematically found to be lower in solutions with lower DO, lower HCO3− concentrations and longer immersion time. The SEM analyses reveal that the iron dissolution rate was more severe in solutions containing higher DO. The decrease in corrosion rate can be attributed to the formation of a passive layer containing mainly α -FeO (OH and ( γ -Fe2O3/Fe3O4 as confirmed by the X-ray diffractometry (XRD and X-ray photoelectron spectroscopy (XPS. Passivation of mild steel is evident in electrochemical test at ≈ −600 mVSCE at pH ≥ 8 in dearated ( ≤ 0.8 ppm DO chloride bicarbonate solution under dynamic conditions.

  16. Development Of ABEC Column For Separation Of Tc-99 From Northstar Dissolved Target Solution

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Dominique C. [Argonne National Lab. (ANL), Argonne, IL (United States); Bennett, Megan E. [Argonne National Lab. (ANL), Argonne, IL (United States); Naik, Seema R. [Argonne National Lab. (ANL), Argonne, IL (United States); ling, lei [Argonne National Lab. (ANL), Argonne, IL (United States); Wang, N-H. Linda [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-01

    Batch and column breakthrough experiments were performed to determine isotherms and mass-transfer parameters for adsorption of Tc on aqueous biphasic extraction chromatographic (ABEC) sorbent in two solutions: 200 g/L Mo, 5.1 M K+, 1 M OH-, and 0.1 M NO3- (Solution A) and 200 g/L Mo, 9.3 M K+, 5 M OH-, and 0.1 M NO3- (Solution B). Good agreement was found between the isotherm values obtained by batch and column breakthrough studies for both Solutions A and B. Potassium-pertechnetate intra-particle diffusivity on ABEC resin was estimated by VERSE simulations, and good agreement was found among a series of column-breakthrough experiments at varying flow velocities, column sizes, and technetium concentrations. However, testing of 10 cc cartridges provided by NorthStar with Solutions A and B did not give satisfactory results, as significant Tc breakthrough was observed and ABEC cartridge performance varied widely among experiments. These different experimental results are believed to be due to inconsistent preparation of the ABEC resin prior to packing and/or inconsistent packing.

  17. Study on safety of crystallization method applied to dissolver solution in fast breeder reactor reprocessing

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Fujine, Yukio; Asakura, Toshihide; Murazaki, Minoru; Koyama, Tomozo; Sakakibara, Tetsuro; Shibata, Atsuhiro

    1999-03-01

    The crystallization method is proposed to apply for recovery of uranium from dissolution liquid, enabling to reduce handling materials in later stages of reprocessing used fast breeder reactor (FBR) fuels. This report studies possible safety problems accompanied by the proposed method. Crystallization process was first defined in the whole reprocessing process, and the quantity and the kind of treated fuel were specified. Possible problems, such as criticality, shielding, fire/explosion, and confinement, were then investigated; and the events that might induce accidental incidents were discussed. Criticality, above all the incidents, was further studied by considering exampled criticality control of the crystallization process. For crystallization equipment, in particular, evaluation models were set up in normal and accidental operation conditions. Related data were selected out from the nuclear criticality safety handbooks. The theoretical densities of plutonium nitrates, which give basic and important information, were estimated in this report based on the crystal structure data. The criticality limit of crystallization equipment was calculated based on the above information. (author)

  18. Effects of solid fission products forming dissolved oxide (Nd) and metallic precipitate (Ru) on the thermal conductivity of uranium base oxide fuel

    International Nuclear Information System (INIS)

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

    2007-01-01

    The effects of solid fission products on the thermal conductivity of uranium base oxide nuclear fuel were experimentally investigated. Neodymium (Nd) and ruthenium (Ru) were added to represent the physical states of solid fission products such as 'dissolved oxide' and 'metallic precipitate', respectively. Thermal conductivity was determined on the basis of the thermal diffusivity, density and specific heat values. The effects of the additives on the thermal conductivity were quantified in the form of the thermal resistivity equation - the reciprocal of the phonon conduction equation - which was determined from the measured data. It is concluded that the thermal conductivity of the irradiated nuclear fuel is affected by both the 'dissolved oxide' and the 'metallic precipitate', however, the effects are in the opposite direction and the 'dissolved oxide' influences the thermal conductivity more significantly than that of the 'metallic precipitate'

  19. Massachusetts Fuel Cell Bus Project: Demonstrating a Total Transit Solution for Fuel Cell Electric Buses in Boston

    Energy Technology Data Exchange (ETDEWEB)

    2017-05-22

    The Federal Transit Administration's National Fuel Cell Bus Program focuses on developing commercially viable fuel cell bus technologies. Nuvera is leading the Massachusetts Fuel Cell Bus project to demonstrate a complete transit solution for fuel cell electric buses that includes one bus and an on-site hydrogen generation station for the Massachusetts Bay Transportation Authority (MBTA). A team consisting of ElDorado National, BAE Systems, and Ballard Power Systems built the fuel cell electric bus, and Nuvera is providing its PowerTap on-site hydrogen generator to provide fuel for the bus.

  20. Effects of dissolved iron and chromium on the performance of direct methanol fuel cell

    International Nuclear Information System (INIS)

    Chen, Weimin; Xin, Qin; Sun, Gongquan; Yang, Shaohua; Zhou, Zhenhua; Mao, Qing; Sun, Pichang

    2007-01-01

    Effects of Fe 3+ and Cr 3+ ions on the performance of direct methanol fuel cell were investigated. The results show that the cell performance decreased remarkably when the concentration of Fe 3+ or Cr 3+ exceeded 1 x 10 -4 mol L -1 . Fe 3+ displayed a strong negative effect on the catalytic oxidation of methanol, while Cr 3+ affected the cell performance primarily by exchanging with protons of the membrane/ionomer and resulted in ionic conductivity losses. Complete recovery of the cell performance was not obtained after flushing the cell with deionized water

  1. Synthesis of Diopside by Solution Combustion Process Using Glycine Fuel

    Science.gov (United States)

    Sherikar, Baburao N.; Umarji, A. M.

    Nano ceramic Diopside (CaMgSi2O6) powders are synthesized by Solution Combustion Process(SCS) using Calcium nitrate, Magnesium nitrate as oxidizer and glycine as fuel, fumed silica as silica source. Ammonium nitrate (AN) is used as extra oxidizer. Effect of AN on Diopside phase formation is investigated. The adiabatic flame temperatures are calculated theoretically for varying amount of AN according to thermodynamic concept and correlated with the observed flame temperatures. A “Multi channel thermocouple setup connected to computer interfaced Keithley multi voltmeter 2700” is used to monitor the thermal events during the process. An interpretation based on maximum combustion temperature and the amount of gases produced during reaction for various AN compositions has been proposed for the nature of combustion and its correlation with the characteristics of as synthesized powder. These powders are characterized by XRD, SEM showing that the powders are composed of polycrystalline oxides with crystallite size of 58nm to 74nm.

  2. Research program and uses of the solution fueled reactor SILENE

    International Nuclear Information System (INIS)

    Barbry, F.; Ratel, R.

    1985-09-01

    Designed and operated by the Nuclear Protection and Safety Institute of the CEA, SILENE is an original small sized reactor fueled with an uranyl nitrate solution. The reactor is capable to operate in three modes: ''Pulse'' operation (high power levels up to 1000 Megawatts during several millisecond), ''Free evolution'' operation (simulation of criticality accident excursions), ''Steady state'' operation in a power range of 0.01 W to 1 kW. The core can be surrounded by appropriate shields (lead, polyethylene) to vary the leakage radiations and the gamma to neutron dose ratio. It's possible to insert in the central cavity of the annular core vessel some capsules, devices or samples to be submitted to very high radiations levels. The research activities are mainly devoted towards nuclear safety studies: the criticality accident studies, and the behavior of oxide fuels under transient conditions. Some examples of tests are presented. As to other applications of the SILENE facility, the main studies now in progress deal with: designing and calibration of Health physics intrumentation, neutron and gamma dosimetry, and, radiobiology. Once the characteristics of radiation field are qualified by calculations and experimental techniques, SILENE will be proposed as a reference source [fr

  3. Solutions for reducing dissolved hydrogen sulphide in the Black Sea by electrochemical oxidation

    International Nuclear Information System (INIS)

    Ciocanea, Adrian; Budea, Sanda; Radulescu, Gabriel

    2007-01-01

    Anaerobic disintegration of organic matter is a particular phenomenon in the Black Sea because of the set up of deposits of hydrogen sulphide, H 2 S, having high concentrations. The formation of such deposits is due to the absence of upward streams at depths larger than 100 meters. In Black Sea there is an oxic layer located roughly between 50 and 200 meters from which downwards begins the anoxic layer. If the equilibrium in Black Sea is not kept under control, an ecological disaster is possible. The first signals will be observed in surface waters, than, if the equilibrium is further disturbed the depth sulphides and the hydrogen sulphide deposits can develop up to inflammable and even explosive phases. This paper presents some solutions to reduce the hydrogen sulphide from Black Sea with a particular stress upon the electrochemical method. (authors)

  4. Examination of hydrogen-bonding interactions between dissolved solutes and alkylbenzene solvents based on Abraham model correlations derived from measured enthalpies of solvation

    Energy Technology Data Exchange (ETDEWEB)

    Varfolomeev, Mikhail A.; Rakipov, Ilnaz T. [Chemical Institute, Kazan Federal University, Kremlevskaya 18, Kazan 420008 (Russian Federation); Acree, William E., E-mail: acree@unt.edu [Department of Chemistry, 1155 Union Circle # 305070, University of North Texas, Denton, TX 76203-5017 (United States); Brumfield, Michela [Department of Chemistry, 1155 Union Circle # 305070, University of North Texas, Denton, TX 76203-5017 (United States); Abraham, Michael H. [Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ (United Kingdom)

    2014-10-20

    Highlights: • Enthalpies of solution measured for 48 solutes dissolved in mesitylene. • Enthalpies of solution measured for 81 solutes dissolved in p-xylene. • Abraham model correlations derived for enthalpies of solvation of solutes in mesitylene. • Abraham model correlations derived for enthalpies of solvation of solutes in p-xylene. • Hydrogen-bonding enthalpies reported for interactions of aromatic hydrocarbons with hydrogen-bond acidic solutes. - Abstract: Enthalpies of solution at infinite dilution of 48 organic solutes in mesitylene and 81 organic solutes in p-xylene were measured using isothermal solution calorimeter. Enthalpies of solvation for 92 organic vapors and gaseous solutes in mesitylene and for 130 gaseous compounds in p-xylene were determined from the experimental and literature data. Abraham model correlations are determined from the experimental enthalpy of solvation data. The derived correlations describe the experimental gas-to-mesitylene and gas-to-p-xylene solvation enthalpies to within average standard deviations of 1.87 kJ mol{sup −1} and 2.08 kJ mol{sup −1}, respectively. Enthalpies of X-H⋯π (X-O, N, and C) hydrogen bond formation of proton donor solutes (alcohols, amines, chlorinated hydrocarbons etc.) with mesitylene and p-xylene were calculated based on the Abraham solvation equation. Obtained values are in good agreement with the results determined using conventional methods.

  5. Bio fuel ash in a road construction: impact on soil solution chemistry.

    Science.gov (United States)

    Thurdin, R T; van Hees, P A W; Bylund, D; Lundström, U S

    2006-01-01

    Limited natural resources and landfill space, as well as increasing amounts of ash produced from incineration of bio fuel and municipal solid waste, have created a demand for useful applications of ash, of which road construction is one application. Along national road 90, situated about 20 km west of Sollefteå in the middle of Sweden, an experiment road was constructed with a 40 cm bio fuel ash layer. The environmental impact of the ash layer was evaluated from soil solutions obtained by centrifugation of soil samples taken on four occasions during 2001-2003. Soil samples were taken in the ash layer, below the ash layer at two depths in the road and in the ditch. In the soil solutions, pH, conductivity, dissolved organic carbon (DOC) and the total concentration of cations (metals) and anions were determined. Two years after the application of the ash layers in the test road, the concentrations in the ash layer of K, SO4, Zn, and Hg had increased significantly while the concentration of Se, Mo and Cd had decreased significantly. Below the ash layer in the road an initial increase of pH was observed and the concentrations of K, SO4, Se, Mo and Cd increased significantly, while the concentrations of Cu and Hg decreased significantly in the road and also in the ditch. Cd was the element showing a potential risk of contamination of the groundwater. The concentrations of Ca in the ash layer indicated an ongoing hardening, which is important for the leaching rate and the strength of the road construction.

  6. Extraction of iron(III) with diphenyl-2-pyridylmethane dissolved in benzene from aqueous chloride solutions

    International Nuclear Information System (INIS)

    Suhail Ahmed; Shamas-Ud-Zuha; Abdul Ghafoor; Ejaz, M.

    1978-01-01

    The mechanism of extraction has been investigated by partition, slope analysis and loading-ratio data. The results obtained give a picture of the mechanism of extraction of FeCl 4 - ions in relation to the hydration and solvation of the compound extracted. The possible formula of the extracted species is (DPPM)sub(3)Hsub(3)Osup(+)(Hsub(2)O)sub(n)-FeClsub(4)sup(-). In dilute aqueous hydrochloric acid systems the influence of the concentration of a number of salts with cations of different electrical potentials (Ze/r), on iron(III) extraction, has been studied. Splitting of the organic phases occurs at high acid and/or high salt concentrations. The phenomenon is explained on the basis of the variability of the hydration number. Investigations have been made to understand the parameters controlling the extraction of the metal and it is shown that the extraction offers a simple, fast and selective separation method of iron from solutions. (author)

  7. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bohnenstingl, J.; Djoa, S. H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-04-15

    This paper describes a process developed for the retainment and separation of volatile (3H, 129 +131I) and gaseous (85Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 deg K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 deg K and low subpressure; deposition of krypton in solid form at 80 deg K after compression to about 6 bar; decontamination of 85krypton-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, e.g., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1/3 of the full capacity and can treat about 1 m3 STP/h helium, corresponding to a quantity of about 10,000 MW(e) HTGR-fuel reprocessing plant.

  8. Separation of the fission product noble gases krypton and xenon from dissolver off-gas in reprocessing HTGR-fuel

    International Nuclear Information System (INIS)

    Bohnenstingl, J.; Djoa, S.H.; Laser, M.; Mastera, S.; Merz, E.; Morschl, P.

    1976-01-01

    This paper describes a process developed for the retainment and separation of volatile ( 3 H, 129+131 I) and gaseous ( 85 Kr, Xe) fission products from the off-gas produced during dissolution of HTGR-fuel. To prevent unnecessary dilution of liberated noble gases by surrounding atmosphere, a helium purge-gas cycle is applied to enable a coarse fractionating of krypton and xenon by cold-trapping at about 80 0 K after precleaning the gas stream. The process consists of the following steps: deposition of droplets and solid aerosols; chemisorption of iodine on silver impregnated silica gel; catalytic removal of nitrogen oxides and oxygen; drying of the process gas stream; final filtering of abraded solids; deposition of xenon in solid form at 80 0 K and low subpressure; deposition of krypton in solid form at 80 0 K after compression to about 6 bar; decontamination of 85 Kr-containing xenon by batch distillation for eventual industrial utilization; and removal of nitrogen and argon enrichment during continuous operation in the purge-gas stream by inleaking air with charcoal. A continuously operating dissolver vessel, closed to the surrounding atmosphere, yields a very high content of noble gases, i.e., 0.35 vol % krypton and 2.0 vol % xenon. The presented off-gas treatment unit is operated in cold runs with 1 / 3 of the full capacity and can treat about 1 m 3 STP/h helium, corresponding to a quantity of about 10,000 MW/sub e/ HTGR-fuel reprocessing plant

  9. Disposal of Kr-85 separated from the dissolver off-gas of a reprocessing plant for LWR fuels

    International Nuclear Information System (INIS)

    Nommensen, O.

    1981-08-01

    The principle of the radiation protection to keep the radiation load of the population as low as possible requires the development of methods for retaining the radionuclide Krypton 85 seperated off the dissolver waste gas of future reprocessing plants for LWR-nuclear fuel elements. In a recommendation of the RSK the long-termed storage of the Kr-85 in a pressure gas bottle and the marine disposal we considered to be disposal methods low in risk. The present work develops a concept for both of the disposal methods and demonstrates their technical feasibility. The comparison of the cost estimations effected for both of the disposal methods shows that the costs related with the marine disposal of the pressure gas bottles amounting to 1.90 DM/kg of reprocessed U fall by the factor 10 below the costs that result from the surface storage of the bottles. In both cases was referred to a reprocessing capacity of 1400 t U/a corresponding to 50 GW installed nuclear power, thereby accumulating approximately 629 PBq (17 MCi) Kr-85 per year. Both concepts project the seperated radioactive inert gas to be filled in pressure gas bottles in a low temperature rectification plant. Each of the 85 bottles to be filled per year contains 7.4 PBq (200 kCi) Kr-85. (orig./HP) [de

  10. Solar fuels via artificial photosynthesis: From homogeneous photocatalysis in solution to a photoelectrochemical cell

    NARCIS (Netherlands)

    Chen, H.-C.

    2016-01-01

    The conversion and storage of solar energy into fuels provides a valuable solution for the future energy demand of our society. Making fuels via artificial photosynthesis, the so-called solar-to-fuel approach, is viewed as one of the most promising ways to produce clean and renewable energy.

  11. Composition characteristics and regularities of dissolving of hydroxyapatite materials obtained in water solutions with varied content of silicate ions

    Science.gov (United States)

    Solonenko, A. P.

    2018-01-01

    Research aimed at developing new bioactive materials for the repair of defects in bone tissues, do not lose relevance due to the strengthening of the regenerative approach in medicine. From this point of view, materials based on calcium phosphates, including silicate ions, consider as one of the most promising group of substances. Methods of synthesis and properties of hydroxyapatite doped with various amounts of SiO4 4- ions are described in literature. In the present work synthesis of a solid phase in the systems Ca(NO3)2 - (NH4)2HPO4 - Na2SiO3 - NH4OH - H2O (Cca/CP = 1.70) performed with a wide range of sodium silicate additive concentration (y = CSi/CP = 0 ÷ 5). It is established that under the studied conditions at y ≥ 0.3 highly dispersed poorly crystallized apatite containing isomorphic impurities of CO3 2- and SiO4 4- precipitates in a mixture with calcium hydrosilicate and SiO2. It is shown that the resulting composites can gradually dissolve in physiological solution and initiate passive formation of the mineral component of hard tissues.

  12. Silver iodide reduction in aqueous solution: application to iodine enhanced separation during spent nuclear fuels reprocessing

    International Nuclear Information System (INIS)

    Badie, Jerome

    2002-01-01

    Silver iodide is a key-compound in nuclear chemistry either in accidental conditions or during the reprocessing of spent nuclear fuel. In that case, the major part of iodine is released in molecular form into the gaseous phase at the time of dissolution in nitric acid. In French reprocessing plants, iodine is trapped in the dissolver off-gas treatment unit by two successive steps: the first consists in absorption by scrubbing with a caustic soda solution and in the second, residual iodine is removed from the gaseous stream before the stack by chemisorption on mineral porous traps made up of beds of amorphous silica or alumina porous balls impregnated with silver nitrate. Reactions of iodine species with the impregnant are assumed to lead to silver iodide and silver iodate. Enhanced separation policy would make necessary to recover iodine from the filters by silver iodide dissolution during a reducing treatment. After a brief silver-iodine chemical bibliographic review, the possible reagents listed in the literature were studied. The choice has been made to use ascorbic acid and hydroxylamine. An experimental work on silver iodide reduction by this two compounds allowed us to determinate reaction products, stoichiometry and kinetics parameters. Finally, the process has been initiated on stable iodine loaded filters samples. (author) [fr

  13. Improved accountability method for measuring enriched uranium in H-Canyon dissolver solution at the Savannah River Site

    International Nuclear Information System (INIS)

    Maxwell, S.L. III; Satkowski, J.; Mahannah, R.N.

    1992-01-01

    At the Savannah River Site (SRS), accountability measurement of enriched uranium dissolved in H-Canyon is performed using isotope dilution mass spectrometry (IDMS). In the IDMS analytical method, a known quantity of uranium 233 is added to the sample solution containing enriched uranium and fission products. The resulting uranium mixture must first be purified using a separation technique in the shielded analytical(''hot'') cells to lower radioactivity levels by removing fission products. Following this purification, the sample is analyzed by mass spectrometry to determine the total uranium content and isotopic abundance. The magnitude of the response of each uranium isotope in the sample solution and the response of the U 233 spike is measured. By ratioing these responses, relative to the known quantity of the U 233 spike, the uranium content can be determined. A hexane solvent extraction technique, used for years at SRS to remove fission products prior to the mass spectrometry analysis of uranium, has several problems. The hexone method is tedious, requires additional sample clean-up after the purified sample is removed from the shielded cells and requires the use of Resource Conservation and Recovery Act (RCRA)-listed hazardous materials (hexone and chromium compounds). A new high speed separation method that enables a rapid removal of fission products in a shielded cells environment has been developed by the SRS Central Laboratory to replace the hexone method. The new high speed column extraction chromatography technique employs applied vacuum and columns containing tri (2-ethyl-hexyl) phosphate (TEHP) solvent coated on a small particle inert support (SM-7 Bio Beads). The new separation is rapid, user friendly, eliminates the use of the RCA-listed hazardous chemicals and reduces the amount of solid waste generated by the separation method. 2 tabs. 4 figs

  14. Overview of the EBFGT installation solutions applicable for flue gases from various fuels combustion

    International Nuclear Information System (INIS)

    Chmielewski, A.G.; Tyminski, B.; Pawelec, A.; Zimek, Z.; Licki, J.

    2011-01-01

    The overview of the solutions used in EBFGT process and adaptation of process parameters for flue gas from combustion of various fuels was presented. The inlets parameters of flue gas from four fuels with high emission of pollutants, process parameters and process constrain were analysed. Also the main problems of this technology and their solutions were presented. (author)

  15. Overview of the EBFGT installation solutions applicable for flue gases from various fuels combustion

    Energy Technology Data Exchange (ETDEWEB)

    Chmielewski, A. G.; Tyminski, B.; Pawelec, A.; Zimek, Z. [Institute of Nuclear Chemistry and Technology, Warsaw (Poland); Licki, J. [Institute of Atomic Energy, Otwock-Świerk (Poland)

    2011-07-01

    The overview of the solutions used in EBFGT process and adaptation of process parameters for flue gas from combustion of various fuels was presented. The inlets parameters of flue gas from four fuels with high emission of pollutants, process parameters and process constrain were analysed. Also the main problems of this technology and their solutions were presented. (author)

  16. Fuel mix electricity 2020, inventory, problems and solutions

    International Nuclear Information System (INIS)

    Seebregts, A.J.; Snoep, H.J.M.; Van Deurzen, J.; Lensink, S.M.; Van der Welle, A.J.; Wetzels, W.

    2010-04-01

    ECN made an inventory of the fuel mix of the electricity generation in the Netherlands for the year 2020. The inventory is derived from the updated Reference Projections that are based on the Global Economy scenario. This scenario has a relatively high growth of the domestic electricity demand (156 TWh in 2020 compared to about 118 TWh in 2008). Besides the factual inventory, ECN made a quickscan of potential problems associated with high penetration of electricity from wind energy, up to 12,000 MW in 2020. The conclusion is that the so-called 'offpeak hours issue' (in Dutch: 'het daluren issue') is a real potential problem under such scenario assumptions. In case of full wind availability, the total generating capacity consisting of must-run capacity and typical base load power plants with low variable cost of production (nuclear and coal) may exceed the electricity demand (domestic plus net export) in part of these off-peak situations, e.g. during nighttime. The must-run capacity consists, among others, of part of the decentralised CHP installations and waste incinerators. Potential solutions to this 'off-peak hours issue' are: (1) Flexibility in electricity demand ('demand response'); (2) Additional interconnection with neighbouring countries and appropriate market design rules; (3) Storage of electricity; (4) Flexibility of conventional (fossil) supply; (5) Flexibility of the intermittent renewable generation itself; (6) Intelligent grids ('Smart Grids'). Additional detailed analyses and research are needed to address the magnitude of the problem and to analyse the contribution of the various solutions. Such an analysis can provide an indication of an optimal and feasible combination of the solutions identified. Relevant issues are: (a) How often, with which magnitude and under which circumstances will the problem occur?; (b) What will be the effect on the curtailing of wind energy, and is curtailing plausible given current market rules and renewable energy

  17. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    International Nuclear Information System (INIS)

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-01-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel

  18. Removal of triazine-based pollutants from water by carbon nanotubes: Impact of dissolved organic matter (DOM) and solution chemistry.

    Science.gov (United States)

    Engel, Maya; Chefetz, Benny

    2016-12-01

    Adsorption of organic pollutants by carbon nanotubes (CNTs) in the environment or removal of pollutants during water purification require deep understanding of the impacts of the presence of dissolved organic matter (DOM). DOM is an integral part of environmental systems and plays a key role affecting the behavior of organic pollutants. In this study, the effects of solution chemistry (pH and ionic strength) and the presence of DOM on the removal of atrazine and lamotrigine by single-walled CNTs (SWCNTs) was investigated. The solubility of atrazine slightly decreased (∼5%) in the presence of DOM, whereas that of lamotrigine was significantly enhanced (by up to ∼70%). Simultaneous introduction of DOM and pollutant resulted in suppression of removal of both atrazine and lamotrigine, which was attributed to DOM-pollutant competition or blockage of adsorption sites by DOM. However the decrease in removal of lamotrigine was also a result of its complexation with DOM. Pre-introduction of DOM significantly reduced pollutant adsorption by the SWCNTs, whereas introduction of DOM after the pollutant resulted in the release of adsorbed atrazine and lamotrigine from the SWCNTs. These data imply that DOM exhibits higher affinity for the adsorption sites than the triazine-based pollutants. In the absence of DOM atrazine was a more effective competitor than lamotrigine for adsorption sites in SWCNTs. However, competition between pollutants in the presence of DOM revealed lamotrigine as the better competitor. Our findings help unravel the complex DOM-organic pollutant-CNT system and will aid in CNT-implementation in water-purification technologies. Copyright © 2016 Elsevier Ltd. All rights reserved.

  19. Beyond temperature: Clumped isotope signatures in dissolved inorganic carbon species and the influence of solution chemistry on carbonate mineral composition

    Science.gov (United States)

    Tripati, Aradhna K.; Hill, Pamela S.; Eagle, Robert A.; Mosenfelder, Jed L.; Tang, Jianwu; Schauble, Edwin A.; Eiler, John M.; Zeebe, Richard E.; Uchikawa, Joji; Coplen, Tyler B.; Ries, Justin B.; Henry, Drew

    2015-01-01

    “Clumped-isotope” thermometry is an emerging tool to probe the temperature history of surface and subsurface environments based on measurements of the proportion of 13C and 18O isotopes bound to each other within carbonate minerals in 13C18O16O22- groups (heavy isotope “clumps”). Although most clumped isotope geothermometry implicitly presumes carbonate crystals have attained lattice equilibrium (i.e., thermodynamic equilibrium for a mineral, which is independent of solution chemistry), several factors other than temperature, including dissolved inorganic carbon (DIC) speciation may influence mineral isotopic signatures. Therefore we used a combination of approaches to understand the potential influence of different variables on the clumped isotope (and oxygen isotope) composition of minerals.We conducted witherite precipitation experiments at a single temperature and at varied pH to empirically determine 13C-18O bond ordering (Δ47) and δ18O of CO32- and HCO3- molecules at a 25 °C equilibrium. Ab initio cluster models based on density functional theory were used to predict equilibrium 13C-18O bond abundances and δ18O of different DIC species and minerals as a function of temperature. Experiments and theory indicate Δ47 and δ18O compositions of CO32- and HCO3- ions are significantly different from each other. Experiments constrain the Δ47-δ18O slope for a pH effect (0.011 ± 0.001; 12 ⩾ pH ⩾ 7). Rapidly-growing temperate corals exhibit disequilibrium mineral isotopic signatures with a Δ47-δ18O slope of 0.011 ± 0.003, consistent with a pH effect.Our theoretical calculations for carbonate minerals indicate equilibrium lattice calcite values for Δ47 and δ18O are intermediate between HCO3− and CO32−. We analyzed synthetic calcites grown at temperatures ranging from 0.5 to 50 °C with and without the enzyme carbonic anhydrase present. This enzyme catalyzes oxygen isotopic exchange between DIC species and is present in many

  20. Beyond temperature: Clumped isotope signatures in dissolved inorganic carbon species and the influence of solution chemistry on carbonate mineral composition

    Science.gov (United States)

    Tripati, Aradhna K.; Hill, Pamela S.; Eagle, Robert A.; Mosenfelder, Jed L.; Tang, Jianwu; Schauble, Edwin A.; Eiler, John M.; Zeebe, Richard E.; Uchikawa, Joji; Coplen, Tyler B.; Ries, Justin B.; Henry, Drew

    2015-10-01

    ;Clumped-isotope; thermometry is an emerging tool to probe the temperature history of surface and subsurface environments based on measurements of the proportion of 13C and 18O isotopes bound to each other within carbonate minerals in 13C18O16O22- groups (heavy isotope ;clumps;). Although most clumped isotope geothermometry implicitly presumes carbonate crystals have attained lattice equilibrium (i.e., thermodynamic equilibrium for a mineral, which is independent of solution chemistry), several factors other than temperature, including dissolved inorganic carbon (DIC) speciation may influence mineral isotopic signatures. Therefore we used a combination of approaches to understand the potential influence of different variables on the clumped isotope (and oxygen isotope) composition of minerals. We conducted witherite precipitation experiments at a single temperature and at varied pH to empirically determine 13C-18O bond ordering (Δ47) and δ18O of CO32- and HCO3- molecules at a 25 °C equilibrium. Ab initio cluster models based on density functional theory were used to predict equilibrium 13C-18O bond abundances and δ18O of different DIC species and minerals as a function of temperature. Experiments and theory indicate Δ47 and δ18O compositions of CO32- and HCO3- ions are significantly different from each other. Experiments constrain the Δ47-δ18O slope for a pH effect (0.011 ± 0.001; 12 ⩾ pH ⩾ 7). Rapidly-growing temperate corals exhibit disequilibrium mineral isotopic signatures with a Δ47-δ18O slope of 0.011 ± 0.003, consistent with a pH effect. Our theoretical calculations for carbonate minerals indicate equilibrium lattice calcite values for Δ47 and δ18O are intermediate between HCO3- and CO32-. We analyzed synthetic calcites grown at temperatures ranging from 0.5 to 50 °C with and without the enzyme carbonic anhydrase present. This enzyme catalyzes oxygen isotopic exchange between DIC species and is present in many natural systems. The two

  1. Lowering of the critical concentration for micelle formation in aqueous soap solutions by action of truly dissolved hydrocarbon at various temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Markina, Z.N.; Kostova, N.Z.; Rebinder, P.A.

    1970-03-01

    The effect of dissolved hydrocarbons (octane, benzene, and ethylbenzene) on critical micelle concentration of aqueous solutions of sodium salts of fatty acids from caproate to sodium myristate at various temperatures was studied. Experimental results showed that formation of micelles is promoted by presence of hydrocarbons dissolved in the water phase. Such solutions have below normal critical micelle concentration. The change in critical micelle concentration decreases with increase in length of hydrocarbon chain in the soap molecule and with decrease of hydrocarbon solubility in pure water. The nature of the hydrocarbon also affects the forms and dimension of the micelle. Aromatic hydrocarbons increase micelle volume and greatly decrease C.M.C., while aliphatic hydrocarbons decrease C.M.C. slightly. (12 refs.)

  2. Parameters affecting level measurement interpretation of nuclear fuel solutions

    International Nuclear Information System (INIS)

    Hunt, B.A.; Landat, D.A.

    1999-01-01

    This paper describes a level measurement technique commonly used in the measurement of radioactive liquids and equipment utilised by the inspectors for safeguards purposes. Some of the influencing parameters affecting the measurement results by this technique are characterised. An essential requisite for successful process operations in chemical facilities involving liquids generally require some physical measurements to be made in-line for both process and quality control in order to achieve the necessary final product specifications . In nuclear fuel reprocessing facilities, the same objectives apply coupled however with an additional requirement of achieving nuclear material accountancy and control. In view of the strategic importance of some of the process vessels in nuclear facilities, accountancy has to be supported by volume and density measurements of low uncertainty. Inspectors therefore require instruments which are at the very least as good as or better than operator's equipment. The classical measurement technique and most widely applied for process liquids in nuclear installations is the bubbler probe or dip-tube technique. Here a regulated flow of air passes through tubes inserted to various depths into the vessel and pressure readings are measured which are a function of the presence of liquid height and density of solution in the tank. These readings, taken together with a pre-determined calibration curve are sufficient for the volume and amount of liquor in a tank to be quantified. All measurement equipment and instrumentation are long distances from the tank environment. The key physical parameter to measure at this location is therefore pressure. Equipment designed developed, commissioned and tested in the tank measurement facilities at Ispra and in nuclear installations in Europe, Japan and the USA, house digital pressure transducer modules with manufacture's declared features of better than 0.01% accuracy and long term stability of 0.01% full

  3. [Effects of forest regeneration patterns on the quantity and chemical structure of soil solution dissolved organic matter in a subtropical forest.

    Science.gov (United States)

    Yuan, Xiao Chun; Lin, Wei Sheng; Pu, Xiao Ting; Yang, Zhi Rong; Zheng, Wei; Chen, Yue Min; Yang, Yu Sheng

    2016-06-01

    Using the negative pressure sampling method, the concentrations and spectral characte-ristics of dissolved organic matter (DOM) of soil solution were studied at 0-15, 15-30, 30-60 cm layers in Castanopsis carlesii forest (BF), human-assisted naturally regenerated C. carlesii forest (RF), C. carlesii plantation (CP) in evergreen broad-leaved forests in Sanming City, Fujian Pro-vince. The results showed that the overall trend of dissolved organic carbon (DOC) concentrations in soil solution was RF>CP>BF, and the concentration of dissolved organic nitrogen (DON) was highest in C. carlesii plantation. The concentrations of DOC and DON in surface soil (0-15 cm) were all significantly higher than in the subsurface (30-60 cm). The aromatic index (AI) was in the order of RF>CP>BF, and as a whole, the highest AI was observed in the surface soil. Higher fluorescence intensity and a short wave absorption peak (320 nm) were observed in C. carlesii plantation, suggesting the surface soil of C. carlesii plantation was rich in decomposed substance content, while the degree of humification was lower. A medium wave absorption peak (380 nm) was observed in human-assisted naturally regenerated C. carlesii forest, indicating the degree of humification was higher which would contribute to the storage of soil fertility. In addition, DOM characte-ristics in 30-60 cm soil solution were almost unaffected by forest regeneration patterns.

  4. ENVIRONMENTAL TECHNOLOGY VERIFICATION REPORT, JCH FUEL SOLUTIONS, INC., JCH ENVIRO AUTOMATED FUEL CLEANING AND MAINTENANCE SYSTEM

    Science.gov (United States)

    The verification testing was conducted at the Cl facility in North Las Vegas, NV, on July 17 and 18, 2001. During this period, engine emissions, fuel consumption, and fuel quality were evaluated with contaminated and cleaned fuel.To facilitate this verification, JCH repre...

  5. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    Science.gov (United States)

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  6. [Effects of nitrogen deposition on the concentration and spectral characteristics of dissolved organic matter in soil solution in a young Cunninghamia lanceolata plantation.

    Science.gov (United States)

    Yuan, Xiao Chun; Chen, Yue Min; Yuan, Shuo; Zheng, Wei; Si, You Tao; Yuan, Zhi Peng; Lin, Wei Sheng; Yang, Yu Sheng

    2017-01-01

    To study the effects of nitrogen deposition on the concentration and spectral characteristics of dissolved organic matter (DOM) in the forest soil solution from the subtropical Cunninghamia lanceolata plantation, using negative pressure sampling method, the dynamics of DOM in soil solutions from 0-15 and 15-30 cm soil layer was monitored for two years and the spectroscopic features of DOM were analyzed. The results showed that nitrogen deposition significantly reduced the concentration of dissolved organic carbon (DOC), and increased the aromatic index (AI) and the humic index (HIX), but had no significant effect on dissolved organic nitrogen (DON) concentration in both soil layers. There was obvious seasonal variation in DOM concentration of the soil solution, which was prominently higher in summer and autumn than in spring and winter.Fourier-transform infrared (FTIR) absorption spectrometry indicated that the DOM in forest soil solution had absorption peaks in the similar position of six regions, being the highest in wave number of 1145-1149 cm -1 . Three-dimensional fluorescence spectra indicated that DOM was mainly consisted of protein-like substances (Ex/Em=230 nm/300 nm) and microbial degradation products (Ex/Em=275 nm/300 nm). The availability of protein-like substances from 0-15 cm soil layer was reduced in the nitrogen treatments. Nitrogen deposition significantly reduced the concentration of DOC in soil solution, maybe largely by reducing soil pH, inhibiting soil carbon mineralization and stimulating plant growth. In particular, the decline of DOC concentration in the surface layer was due to the production inhibition of the protein-like substances and carboxylic acids. Short-term nitrogen deposition might be beneficial to the maintenance of soil fertility, while the long-term accumulation of nitrogen deposition might lead to the hard utilization of soil nutrients.

  7. Education - path towards solution regarding disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Klein, D.E.

    1991-01-01

    Education, not emotional reaction, is the path to take in the safe disposal of spent nuclear fuel. Education is needed at all levels: Elementary schools, secondary schools, two-year colleges, four-year colleges, graduate schools, and adult education. The Office of Civilian Radioactive Waste Management (OCRWM) should not be expected to tackle this problem alone. Assistance is needed from local communities, schools, and state and federal governments. However, OCRWM can lay the foundation for a comprehensive educational plan directed specifically at educating the public on the spent nuclear fuel issue and OCRWM can begin the implementation of this plan

  8. A complete NUHOMS {sup registered} solution for storage and transport of high burnup spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bondre, J. [Transnuclear, Inc. (AREVA Group), Fremont, CA (United States)

    2004-07-01

    The discharge burnups of spent fuel from nuclear power plants keep increasing with plants discharging or planning to discharge fuel with burnups in excess of 60,000 MWD/MTU. Due to limited capacity of spent fuel pools, transfer of older cooler spent fuel from fuel pool to dry storage, and very limited options for transport of spent fuel, there is a critical need for dry storage of high burnup, higher heat load spent fuel so that plants could maintain their full core offload reserve capability. A typical NUHOMS {sup registered} solution for dry spent fuel storage is shown in the Figure 1. Transnuclear, Inc. offers two advanced NUHOMS {sup registered} solutions for the storage and transportation of high burnup fuel. One includes the NUHOMS {sup registered} 24PTH system for plants with 90.7 Metric Ton (MT) crane capacity; the other offers the higher capacity NUHOMS {sup registered} 32PTH system for higher crane capacity. These systems include NUHOMS {sup registered} - 24PTH and -32PTH Transportable Canisters stored in a concrete storage overpack (HSM-H). These canisters are designed to meet all the requirements of both storage and transport regulations. They are designed to be transported off-site either directly from the spent fuel pool or from the storage overpack in a suitable transport cask.

  9. Surface micro-dissolve method of imparting self-cleaning property to cotton fabrics in NaOH/urea aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Tao; Hu, Ruimin; Zhao, Zhenyun [College of Textile & Garment, Southwest University, 400716, Chongqing (China); Liu, Yiping [College of Textile & Garment, Southwest University, 400716, Chongqing (China); Chongqing Engineering Research Center of Biomaterial Fiber and Modern Textile, 400716, Chongqing (China); Lu, Ming, E-mail: lumingswu@163.com [College of Textile & Garment, Southwest University, 400716, Chongqing (China); Chongqing Engineering Research Center of Biomaterial Fiber and Modern Textile, 400716, Chongqing (China)

    2017-04-01

    Highlights: • A novel micro-dissolved process was carried out to embedding commercial titanium dioxide nanoparticles into cotton fabric with NaOH/urea aqueous solution. • X-ray diffraction pattern of modified fabrics shown that the cellulose structure of modified fabrics had not changed. • Modified cotton fabrics demonstrated favourable photocatalytic self-cleaning performance while tensile strength and whiteness of treated fabrics also expressed an increasement slightly. - Abstract: A simple and economical micro-dissolved process of embedding titanium dioxide (TiO{sub 2}) nanoparticles into surface zone of cotton fabrics was developed. TiO{sub 2} was coated on cotton fabrics in 7% wt NaOH/12% wt urea aqueous solution at low temperature. Photocatalytic efficiency of cotton fabrics treated with TiO{sub 2} nanoparticles was studied upon measuring the photocatalytic decoloration of Rhodamine B (RhB) under ultraviolet irradiation. Self-cleaning property of cotton fabric coated with TiO{sub 2} was evaluated with color depth of samples (K/S value). The treated fabrics were characterized using scanning electron microscopy (SEM), energy dispersive spectrometer (EDS), Fourier transform infrared spectroscopy (FITR), tensile strength, stiffness and whiteness. The results indicated, TiO{sub 2} nanoparticles could be embedded on the surface layer of cotton fabrics throuth surface micro-dissolve method. Treated cotton fabrics possessed distinct photocatalytic efficiency and self-cleaning properties. Tensile strength and whiteness of modified cotton fabrics appeared moderately increasement.

  10. Hydrogen as a renewable and sustainable solution in reducing global fossil fuel consumption

    International Nuclear Information System (INIS)

    Midilli, Adnan; Dincer, Ibrahim

    2008-01-01

    In this paper, hydrogen is considered as a renewable and sustainable solution for reducing global fossil fuel consumption and combating global warming and studied exergetically through a parametric performance analysis. The environmental impact results are then compared with the ones obtained for fossil fuels. In this regard, some exergetic expressions are derived depending primarily upon the exergetic utilization ratios of fossil fuels and hydrogen: the fossil fuel based global waste exergy factor, hydrogen based global exergetic efficiency, fossil fuel based global irreversibility coefficient and hydrogen based global exergetic indicator. These relations incorporate predicted exergetic utilization ratios for hydrogen energy from non-fossil fuel resources such as water, etc., and are used to investigate whether or not exergetic utilization of hydrogen can significantly reduce the fossil fuel based global irreversibility coefficient (ranging from 1 to +∞) indicating the fossil fuel consumption and contribute to increase the hydrogen based global exergetic indicator (ranging from 0 to 1) indicating the hydrogen utilization at a certain ratio of fossil fuel utilization. In order to verify all these exergetic expressions, the actual fossil fuel consumption and production data are taken from the literature. Due to the unavailability of appropriate hydrogen data for analysis, it is assumed that the utilization ratios of hydrogen are ranged between 0 and 1. For the verification of these parameters, the variations of fossil fuel based global irreversibility coefficient and hydrogen based global exergetic indicator as the functions of fossil fuel based global waste exergy factor, hydrogen based global exergetic efficiency and exergetic utilization of hydrogen from non-fossil fuels are analyzed and discussed in detail. Consequently, if exergetic utilization ratio of hydrogen from non-fossil fuel sources at a certain exergetic utilization ratio of fossil fuels increases

  11. Commercial solutions [for dry spent fuel storage casks

    International Nuclear Information System (INIS)

    Howe, W.F.; Pennington, C.W.; Hobbs, J.; Lee, W.; Thomas, B.D.; Dibert, D.J.

    1996-01-01

    In the aftermath of the termination of the DOE's MPC (Multi-Purpose Canister) programme, commercial suppliers are coming forward with new or updated systems to meet utility needs. Leading vendors describe the advantages of their systems for dry spent fuel storage and transport. (Author)

  12. Spent fuel disposal: is the underground the sole solution?

    International Nuclear Information System (INIS)

    Nachmilner, L.

    1997-01-01

    The following 4 major approaches to spent fuel disposal are discussed: permanent storage in an underground repository, reprocessing, partitioning and transmutation, and accelerator driven transmutation. It is concluded that underground disposal will remain the basic option for the near future, although pursuing the other methods is certainly worth while. (P.A.)

  13. Solution of the conjugated heat transfer problem for the fuel elements assemblies

    International Nuclear Information System (INIS)

    Golba, V.S.; Ivanenko, I.J.; Zinina, G.A.

    1997-01-01

    The paper presents the assemblies conjugated heat conductivity problem calculation and experimental method. The method is based on the temperature superposition modified concept and subchannel method and allows to predict the fuel elements surface temperatures with availability of fuel elements inside structure of any complication caused by technological and working defects and with availability of depositions with low heat conductivity on the fuel elements surfaces. According to the method developed the partial solutions of the heat conductivity equation at the heat removal boundaries (solid-liquid) are found separately for the fuel elements and for the liquid. The heat conductivity equation partial solutions for the fuel elements are predicted by calculations. The coolant heat conductivity equation partial solution ('influence functions') data massif is obtained in present work experimentally in the fuel assembly model consists of 7 tube bundle of fuel elements imitators placed in right grating with relative grating step equal to 1.1 and cooled by eutectic alloy Pb-Bi. It is shown that 'subchannel prediction method' decreases the crosswise heat transfer in comparison with crosswise heat transfer, when the fuel element inside structure is taken into account. Also in the paper it is shown that it is possible to realize the assembly temperature prediction method suggested without carrying out the experiments in the assembly's model in order to get the external problem influence functions'. (author)

  14. Natural dissolved organic matter mobilizes Cd but does not affect the Cd uptake by the green algae Pseudokirchneriella subcapitata (Korschikov) in resin buffered solutions

    Energy Technology Data Exchange (ETDEWEB)

    Verheyen, Liesbeth, E-mail: verheyenliesbeth@gmail.com; Versieren, Liske, E-mail: liske.versieren@ees.kuleuven.be; Smolders, Erik, E-mail: erik.smolders@ees.kuleuven.be

    2014-09-15

    Highlights: • Different DOM samples were added to solutions with a resin buffered Cd{sup 2+} activity. • This increased total dissolved Cd by factors 3–16 due to complexation reactions. • Cd uptake in algae was unaffected or increased maximally 1.6 fold upon addition. • Free Cd{sup 2+} is the main bioavailable form of Cd for algae in well buffered solutions. - Abstract: Natural dissolved organic matter (DOM) can have contrasting effects on metal bioaccumulation in algae because of complexation reactions that reduce free metal ion concentrations and because of DOM adsorption to algal surfaces which promote metal adsorption. This study was set up to reveal the role of different natural DOM samples on cadmium (Cd) uptake by the green algae Pseudokirchneriella subcapitata (Korschikov). Six different DOM samples were collected from natural freshwater systems and isolated by reverse osmosis. In addition, one {sup 13}C enriched DOM sample was isolated from soil to trace DOM adsorption to algae. Algae were exposed to standardized solutions with or without these DOM samples, each exposed at equal DOM concentrations and at equal non-toxic Cd{sup 2+} activity (∼4 nM) that was buffered with a resin. The DOM increased total dissolved Cd by factors 3–16 due to complexation reactions at equal Cd{sup 2+} activity. In contrast, the Cd uptake was unaffected by DOM or increased maximally 1.6 fold ({sup 13}C enriched DOM). The {sup 13}C analysis revealed that maximally 6% of algal C was derived from DOM and that this can explain the small increase in biomass Cd. It is concluded that free Cd{sup 2+} and not DOM-complexed Cd is the main bioavailable form of Cd when solution Cd{sup 2+} is well buffered.

  15. Distributions of 14 elements on 60 selected absorbers from two simulant solutions (acid-dissolved sludge and alkaline supernate) for Hanford HLW Tank 102-SY

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1993-10-01

    Sixty commercially available or experimental absorber materials were evaluated for partitioning high-level radioactive waste. These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. The distributions of 14 elements onto each absorber were measured from simulated solutions that represent acid-dissolved sludge and alkaline supernate solutions from Hanford high-level waste (HLW) Tank 102-SY. The selected elements, which represent fission products (Ce, Cs, Sr, Tc, and Y); actinides (U, Pu, and Am); and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr), were traced by radionuclides and assayed by gamma spectrometry. Distribution coefficients for each of the 1680 element/absorber/solution combinations were measured for dynamic contact periods of 30 min, 2 h, and 6 h to provide sorption kinetics information for the specified elements from these complex media. More than 5000 measured distribution coefficients are tabulated

  16. A review of irradiation induced re-solution in oxide fuels

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1980-01-01

    The paper reviews the existing experimental evidence for irradiation induced re-solution and also possible explanations for the mechanism. The importance of re-solution is considered with regard to intragranular bubbles and the accumulation of gas on grain boundaries. It is concluded that re-solution is most effective at low temperatures and could account for the present concern over gas release in high burn-up water reactor fuel assemblies. (author)

  17. Analysis of dissolved benzene plumes and methyl tertiary butyl ether (MTBE) plumes in ground water at leaking underground fuel tank (LUFT) sites

    International Nuclear Information System (INIS)

    Happel, A.M.; Rice, D.; Beckenbach, E.; Savalin, L.; Temko, H.; Rempel, R.; Dooher, B.

    1996-11-01

    The 1990 Clean Air Act Amendments mandate the addition of oxygenates to gasoline products to abate air pollution. Currently, many areas of the country utilize oxygenated or reformulated fuel containing 15- percent and I I-percent MTBE by volume, respectively. This increased use of MTBE in gasoline products has resulted in accidental point source releases of MTBE containing gasoline products to ground water. Recent studies have shown MTBE to be frequently detected in samples of shallow ground water from urban areas throughout the United States (Squillace et al., 1995). Knowledge of the subsurface fate and transport of MTBE in ground water at leaking underground fuel tank (LUFT) sites and the spatial extent of MTBE plumes is needed to address these releases. The goal of this research is to utilize data from a large number of LUFT sites to gain insights into the fate, transport, and spatial extent of MTBE plumes. Specific goals include defining the spatial configuration of dissolved MTBE plumes, evaluating plume stability or degradation over time, evaluating the impact of point source releases of MTBE to ground water, and attempting to identify the controlling factors influencing the magnitude and extent of the MTBE plumes. We are examining the relationships between dissolved TPH, BTEX, and MTBE plumes at LUFT sites using parallel approaches of best professional judgment and a computer-aided plume model fitting procedure to determine plume parameters. Here we present our initial results comparing dissolved benzene and MTBE plumes lengths, the statistical significance of these results, and configuration of benzene and MTBE plumes at individual LUFT sites

  18. Taxonomy of Means and Ends in Aquaculture Production—Part 2: The Technical Solutions of Controlling Solids, Dissolved Gasses and pH

    Directory of Open Access Journals (Sweden)

    Bjorgvin Vilbergsson

    2016-09-01

    Full Text Available In engineering design, knowing the relationship between the means (technique and the end (desired function or outcome is essential. The means in Aquaculture are technical solutions like airlifts that are used to achive desired functionality (an end like controlling dissolved gasses. In previous work, the authors identified possible functions by viewing aquaculture production systems as transformation processes in which inputs are transformed by treatment techniques (means and produce outputs (ends. The current work creates an overview of technical solutions of treatment functions for both design and research purposes. A comprehensive literature review of all areas of technical solutions is identified and categorized into a visual taxonomy of the treatment functions for controlling solids, controlling dissolved gasses and controlling pH alkalinity and hardness. This article is the second in a sequence of four and partly presents the treatments functions in the taxonomy. The other articles in this series present complementary aspects of this research: Part 1, A transformational view on aquaculture and functions divided into input, treatment and output functions; Part 2, The current taxonomy paper; Part 3, The second part of the taxonomy; and Part 4, Mapping of the means (techniques for multiple treatment functions.

  19. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    International Nuclear Information System (INIS)

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

    1997-10-01

    The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O 2 F 2 solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs

  20. Aerosol and iodine removal system for the dissolver off-gas in a large fuel reprocessing plant

    International Nuclear Information System (INIS)

    Furrer, J.; Wilhelm, J.G.; Jannakos, K.

    1979-01-01

    A newly developed filter combination for the dissolver off-gas in a reprocessing plant with a throughput of 1400 t/y of heavy metal is presented and single filter components are described. The design principle chosen provides for remote handling and direct disposal in waste drums of 200 l volume. The optimization of housings and filter units is studied on true scale components in the simulated dissolver off-gas of a test facility named PASSAT. This facility will be described. PASSAT will be also used for final testing of the SORPTEX process which is under development. Its concept is included in the paper. The design and function of the new multiway sorption filter providing for complete loading of the iodine sorption material and maintaining continuously high decontamznation factors will also be given. Removal efficiencies measured for aerosols and iodine in an existing reprocessing plant are indicated

  1. Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

    2003-01-01

    In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

  2. Effect of dissolved humic acid on the Pb bioavailability in soil solution and its consequence on ecological risk.

    Science.gov (United States)

    An, Jinsung; Jho, Eun Hea; Nam, Kyoungphile

    2015-04-09

    Current risk characterization in ecological risk assessment does not consider bioavailability of heavy metals, which highly depends on physicochemical properties of environmental media. This study was set to investigate the effect of humic acid (HA), used as a surrogate of organic matter, on Pb toxicity and the subsequent effect on risk characterization in ecological risk assessment. Pb toxicity was assessed using Microtox(®) in the presence and absence of two different forms of HA, particulate HA (pHA) and dissolved HA (dHA). With increasing contact time, the EC10 values increased (i.e., the toxic effects decreased) and the dissolved Pb concentrations of the filtrates decreased. The high correlation (R = 0.88, p < 0.001) between toxic effects determined using both the mixture and its filtrate as exposure media leads us to conclude that the Pb toxicity highly depends on the soluble fraction. Also, reduced Pb toxicity with increasing dHA concentrations, probably due to formation of Pb-dHA complexes, indicated that Pb toxicity largely comes from free Pb ions. Overall, this study shows the effect of HA on metal toxicity alleviation, and emphasizes the need for incorporating the bioavailable heavy metal concentrations in environmental media as a point of exposure in ecological risk assessment. Copyright © 2015 Elsevier B.V. All rights reserved.

  3. Strategies and solutions in the temporary management of spent fuel in Spain

    International Nuclear Information System (INIS)

    Martinez Abad, J. E.; Rivera, M. I.

    2009-01-01

    The basic strategy for the spent fuel and HLW management contemplated in the Sixth General Radioactive Waste Plan focused on the centralised interim storage of spent fuel, based on proved dry storage system technologies, over the time periods required until their definitive or very long term management. Specially, the solution proposed as the most suitable for the Spanish case is the construction of a centralised interim spent fuel and HLW storage facility (ATC) for which as site is being searched. Until it becomes in operation, the interim spent fuel storage will be safety performed in the NPP reracked spent fuel pools or individual ISFSI constructed in the NPP site, in those cases additional storage capacity is required. (Author) 22 refs

  4. Bio-inspired solutions in design for manufacturing of micro fuel cell

    DEFF Research Database (Denmark)

    Omidvarnia, Farzaneh; Hansen, Hans Nørgaard

    2014-01-01

    In this paper the application of biomimetic principles in design for micro manufacturing is investigated. A micro direct methanol fuel cell (μDMFC) for power generation in hearing aid devices is considered as the case study in which the bioinspired functions are replicated. The focus in design of μ......DMFC is mainly on solving the problem of fuel delivery to the anode in the fuel chamber. Two different biological phenomena are suggested, and based on them different bioinspired solutions are proposed and modeled in CAD software. Considering the manufacturing constraints and design specifications...

  5. Post-precipitations from MOX fuel solutions and analysis of microparticle formation in the PUREX process

    International Nuclear Information System (INIS)

    Henkelmann, R.; Baumgaertner, F.; Klein, F.; Niestroj, B.

    1989-01-01

    Subsequent precipitates of feed solutions from reprocessing were examined with the aid of the SEM-EDX method. On the one hand the examinations give information about the particle form and size distribution, on the other hand about the element distribution in single particles with consideration of the radiation data of the fuel. The subsequent precipitation samples which are examined in this study were taken after different residence times of the clarified fuel solutions. The examinations give information about the kind, element frequency, distribution and stoichiometry of single particles of the submicro- and microrange. (RB) [de

  6. Reclamation of zinc-contaminated soil using a dissolved organic carbon solution prepared using liquid fertilizer from food-waste composting.

    Science.gov (United States)

    Chiang, Po-Neng; Tong, Ou-Yang; Chiou, Chyow-San; Lin, Yu-An; Wang, Ming-Kuang; Liu, Cheng-Chung

    2016-01-15

    A liquid fertilizer obtained through food-waste composting can be used for the preparation of a dissolved organic carbon (DOC) solution. In this study, we used the DOC solutions for the remediation of a Zn-contaminated soil (with Zn concentrations up to 992 and 757 mg kg(-1) in topsoil and subsoil, respectively). We then determined the factors that affect Zn removal, such as pH, initial concentration of DOC solution, and washing frequency. Measurements using a Fourier Transform infrared spectrometer (FT-IR) revealed that carboxyl and amide were the major functional groups in the DOC solution obtained from the liquid fertilizer. Two soil washes using 1,500 mg L(-1) DOC solution with a of pH 2.0 at 25°C removed about 43% and 21% of the initial Zn from the topsoil and subsoil, respectively. Following this treatment, the pH of the soil declined from 5.4 to 4.1; organic matter content slightly increased from 6.2 to 6.5%; available ammonium (NH4(+)-N) content increased to 2.4 times the original level; and in the topsoil, the available phosphorus content and the exchangeable potassium content increased by 1.65 and 2.53 times their initial levels, respectively. Copyright © 2015 Elsevier B.V. All rights reserved.

  7. Demonstration of a SANEX Process in Centrifugal Contactors using the CyMe{sub 4}-BTBP Molecule on a Genuine Fuel Solution

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, D.; Christiansen, B.; Glatz, J.P.; Malmbeck, R.; Serrano-Purroy, D. [European Commiss, Joint Res Ctr, Inst Transuranium Elements, D-76125 Karlsruhe, (Germany); Foreman, M.R.S. [Univ Reading, Dept Chem, Reading RG6 6AD, Berks, (United Kingdom); Geist, A. [Forschungszentrum Karlsruhe, Inst Nukl Entsorgung, D-76021 Karlsruhe, (Germany); Modolo, G. [Forschungszentrum Julich, Inst Energy Res Safety Res and Reactor Technol, D-52425 Julich, (Germany); Sorel, C. [Commissariat Energie Atom Valrho, CEA, DRCP SCPS, F-30207 Bagnols Sur Ceze, (France)

    2009-07-01

    Efficient recovery of minor actinides from a genuine spent fuel solution has been successfully demonstrated by the CyMe{sub 4}-BTBP/DMDOHEMA extractant mixture dissolved in octanol. The continuous countercurrent process, in which actinides(III) were separated from lanthanides(III), was carried out in laboratory centrifugal contactors using an optimized flow-sheet involving a total of 16 stages. The process was divided into 9 stages for extraction from a 2 M nitric acid feed solution, 3 stages for lanthanide scrubbing, and 4 stages for actinide back-extraction. Excellent feed decontamination factors for Am (7000) and Cm (1000) were obtained and the recoveries of these elements were higher than 99.9%. More than 99.9% of the lanthanides were directed to the raffinate except Gd for which 0.32% was recovered in the product. (authors)

  8. Solution phase thermodynamics of strong electrolytes based on ionic concentrations, hydration numbers and volumes of dissolved entities

    Czech Academy of Sciences Publication Activity Database

    Heyrovská, Raji

    2013-01-01

    Roč. 24, č. 6 (2013), s. 1895-1901 ISSN 1040-0400 Institutional support: RVO:68081707 Keywords : Solution thermodynamics * Aqueous electrolytes * Partial electrolytic dissociation Subject RIV: BO - Biophysics Impact factor: 1.900, year: 2013

  9. A simple and rapid method for monitoring dissolved oxygen in water with a submersible microbial fuel cell (SBMFC)

    DEFF Research Database (Denmark)

    Zhang, Yifeng; Angelidaki, Irini

    2012-01-01

    Asubmersiblemicrobial fuel cell (SBMFC) was developed as a biosensor forin situand real time monitoring of dissolvedoxygen (DO) in environmental waters. Domestic wastewater was utilized as sole fuel for powering the sensor. The sensor performance was firstly examined with tap water at varying DO...... the sensing activities. The sensor ability was further explored under different environmental conditions (e.g., pH, temperature, conductivity, alternative electron acceptor), and the results indicated that a calibration would be required before field application. Lastly, the sensor was tested with different...

  10. 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

  11. Cost and Fuel Efficient SCR-only Solution for post-2010 HD Emission Standards

    NARCIS (Netherlands)

    Cloudt, R.P.M.; Willems, F.P.T.; Heijden, van der P.

    2009-01-01

    A promising SCR-only solution is presented to meetpost-2010 NOx emission targets for heavy dutyapplications. The proposed concept is based on anengine from a EURO IV SCR application, which isconsidered optimal with respect to fuel economy andcosts. The addition of advanced SCR after

  12. Fuel solution criticality accident studies with the SILENE reactor: phenomenology, consequences and simulated intervention

    International Nuclear Information System (INIS)

    Barbry, F.

    1984-01-01

    After defining the content and the objectives of criticality accident studies, the SILENE reactor, a means of studying fuel solution criticality accidents, is presented. Information obtained from the CRAC and SILENE experimental programs are then presented; they concern power excursion phenomenology, radiological consequences, and finally guide-lines for current and future programs

  13. Cost and fuel efficient SCR-only solution for post-2010 HD emission standards

    NARCIS (Netherlands)

    Cloudt, R.P.M.; Willems, F.P.T.; Heijden, P.V.A.M. van der

    2009-01-01

    A promising SCR-only solution is presented to meet post-2010 NOx emission targets for heavy duty applications. The proposed concept is based on an engine from a EURO IV SCR application, which is considered optimal with respect to fuel economy and costs. The addition of advanced SCR after treatment

  14. Stable Isotopic Composition of Dissolved Organic Nitrogen Fueling Brown Tide in a Semi-Arid Texas Estuary

    Science.gov (United States)

    Campbell, J.; Felix, J. D. D.; Wetz, M.; Cira, E.

    2017-12-01

    Harmful algal blooms (HABs) have the potential to adversely affect the water quality of estuaries and, consequently, their ability to support healthy and diverse ecosystems. Since the early 1990s, Baffin Bay, a semi-arid south Texas estuary, has progressively experienced harmful algal blooms. The primary species of HAB native to the Baffin Bay region, Aureoumbra lagunensis, is unable to utilize nitrate as a nutrient source, but instead relies on forms of reduced nitrogen (such as dissolved organic nitrogen (DON) and ammonium (NH4+)) for survival. DON levels in Baffin Bay (77 ± 10 µM) exceed the DON concentrations of not only typical Texas estuaries, but estuaries worldwide. Additionally, DON accounts for 90% of the total dissolved nitrogen (TDN) in Baffin Bay, followed by NH4+ at 8%, and NO3-+NO2- contributing 2%. Due to the dependence of A. lagunensis on the reduced forms of nitrogen as an energy source and the elevated concentrations of DON throughout the bay, it is important to identify the origin of this nitrogen as well as how it's being processed as it cycles through the ecosystem. The presented work investigates the stable isotopic composition of reactive nitrogen (Nr) (δ15N-DON, δ15N-NH4+, and δ15N-NO3-) in Baffin Bay samples collected monthly at nine stations over the period of one year. The work provides preliminary evidence of Nr sources and mechanisms driving favorable conditions for HAB proliferation. This information can be useful and applicable to estuarine ecosystems in various settings, advancing scientific progress towards mitigating blooms. Additionally, since the elevated concentrations of DON make Baffin Bay uniquely suited to investigate its sources and processing, this project will aid in characterizing the role of this largely unstudied form of Nr, which could provide insight and change perceptions about the role of DON in nitrogen dynamics.

  15. Attractive forces between hydrophobic solid surfaces measured by AFM on the first approach in salt solutions and in the presence of dissolved gases.

    Science.gov (United States)

    Azadi, Mehdi; Nguyen, Anh V; Yakubov, Gleb E

    2015-02-17

    Interfacial gas enrichment of dissolved gases (IGE) has been shown to cover hydrophobic solid surfaces in water. The atomic force microscopy (AFM) data has recently been supported by molecular dynamics simulation. It was demonstrated that IGE is responsible for the unexpected stability and large contact angle of gaseous nanobubbles at the hydrophobic solid-water interface. Here we provide further evidence of the significant effect of IGE on an attractive force between hydrophobic solid surfaces in water. The force in the presence of dissolved gas, i.e., in aerated and nonaerated NaCl solutions (up to 4 M), was measured by the AFM colloidal probe technique. The effect of nanobubble bridging on the attractive force was minimized or eliminated by measuring forces on the first approach of the AFM probe toward the flat hydrophobic surface and by using high salt concentrations to reduce gas solubility. Our results confirm the presence of three types of forces, two of which are long-range attractive forces of capillary bridging origin as caused by either surface nanobubbles or gap-induced cavitation. The third type is a short-range attractive force observed in the absence of interfacial nanobubbles that is attributed to the IGE in the form of a dense gas layer (DGL) at hydrophobic surfaces. Such a force was found to increase with increasing gas saturation and to decrease with decreasing gas solubility.

  16. Solvent wash solution

    International Nuclear Information System (INIS)

    Neace, J.C.

    1986-01-01

    This patent describes a process for removing diluent degradation products from a solvent extraction solution comprising an admixture of an organic extractant for uranium and plutonium and a non-polar organic liquid diluent, which has been used to recover uranium and plutonium from spent nuclear fuel. Comprising combining a wash solution consisting of: (a) water; and (b) a positive amount up to about, an including, 50 volume percent of at least one highly-polar water-miscible organic solvent, based on the total volume of the water and the highly-polar organic solvent, with the solvent extraction solution after uranium and plutonium values have been stripped from the solvent extraction solution, the diluent degradation products dissolving in the highly-polar organic solvent and the extractant and diluent of the extraction solution not dissolving in the highly-polar organic solvent, and separating the highly-polar organic solvent and the extraction solution to obtain a purified extraction solution

  17. Separation of the noble metals ruthenium and palladium from nitric acid solution of the nuclear fuel reprocessing containing complexing agents

    International Nuclear Information System (INIS)

    Ghafourian, H.

    1989-06-01

    Two extraction chromatographic techniques have been developed. N'N diethylthiourea (DETU), which forms complexes with ruthenium that can be retained on an AG50W-X2 ion exchanger, has proved to be a suitable reagent. The structures of these complexes were elucidated by electrophoresis, ion exchange and IR spectroscopy. Under the same conditions Pd forms an insoluble DETU-complex of the formula [Pd(DETU) 4 ] 2+ , which allows the separation of this metal quantitatively. With regard to the application of the developed technique for recovery of the mentioned noble metals from dissolver residues of the nuclear fuel reprocessing, comparative studies were carried out for accompanying fission product nuclides and actinides such as Mo, Tc, Zr, Ce, U and Pu. It was found out that no complex between diethylthiourea and the fission products zirconium, molybdenum and cerium and the actinides uranium, plutonium and americium were formed. Technetium, which was originally present as pertechnetate, is reduced to Tc(IV) and retained on the cation exchanger together with ruthenium. Ruthenium was eluted with 6 M HNO 3 . The efficiency of the developed process has been demonstrated with simulated solutions. The achieved decontamination factors ranged from 10 2 to 10 6 depending on the nuclide. (orig./RB) [de

  18. Dissolved organic carbon in the precipitation of Seoul, Korea: Implications for global wet depositional flux of fossil-fuel derived organic carbon

    Science.gov (United States)

    Yan, Ge; Kim, Guebuem

    2012-11-01

    Precipitation was sampled in Seoul over a one-year period from 2009 to 2010 to investigate the sources and fluxes of atmospheric dissolved organic carbon (DOC). The concentrations of DOC varied from 15 μM to 780 μM, with a volume-weighted average of 94 μM. On the basis of correlation analysis using the commonly acknowledged tracers, such as vanadium, the combustion of fossil-fuels was recognized to be the dominant source. With the aid of air mass backward trajectory analyses, we concluded that the primary fraction of DOC in our precipitation samples originated locally in Korea, albeit the frequent long-range transport from eastern and northeastern China might contribute substantially. In light of the relatively invariant organic carbon to sulfur mass ratios in precipitation over Seoul and other urban regions around the world, the global magnitude of wet depositional DOC originating from fossil-fuels was calculated to be 36 ± 10 Tg C yr-1. Our study further underscores the potentially significant environmental impacts that might be brought about by this anthropogenically derived component of organic carbon in the atmosphere.

  19. Chemical effects induced by dissolving γ-irradiated alkali halides in aqueous nitrate, permanganate and chromate solutions

    International Nuclear Information System (INIS)

    Phansalkar, V.K.; Bapat, L.; Ravishankar, D.

    1982-01-01

    Dissolution of γ-irradiated alkali halides in aqueous solutions of sodium nitrate, potassium permanganate and potassium chromate at neutral pH induces chemical changes leading to the formation of NO 2 - in nitrate, Mn(IV) and Cr(III) species in permanganate and chromate solutions, respectively. Further, the studies on nitrate and permanganate systems show that the amount of NO 2 - and Mn(IV) formed grows by increasing the dose of γ-irradiation of the salt and the amount of irradiated salt. Moreover, the extent of chemical changes effected by irradiated chlorides has been found to be more than that of bromides. The mesh size of the irradiated salt and the presence of scavengers like I - and methanol in the system, affects the yield of NO 2 - . (author)

  20. JASPAS programme task JC-4: Isotopic and isotope dilution analysis of spent fuel solutions by resin bead mass spectrometry

    International Nuclear Information System (INIS)

    Hayashi, N.; Terakado, S.; Kuno, Y.

    1988-05-01

    The use of resin beads for mass spectrometry of U and Pu has been extensively developed at Oak Ridge National Laboratory in the USA and tested in a number of intercomparison experiments between the Safeguards Analytical Laboratory (SAL) of the IAEA and the Power Reactor and Fuel Development Corporation (PNC) - Tokai Reprocessing Plant (TRP) in Japan. Resin beads represent a convenient way to concentrate the U and Pu in spent fuel dissolver solution samples from reprocessing facilities, with the added advantage that fission product elements and other actinides such as Am are removed. Measurements on the resin bead samples at SAL were performed on the ORNL-designed 2-Stage Mass Spectrometer. For the dried tracer samples, the U measurements were obtained on the VG54E instrument and the Pu results were obtained with the Finnigan MAT 261 of SAL. PNC/TRP used a VG54 mass spectrometer and obtained their mass fractionation correction factor for the resin bead measurements from the mixed tracer plus chemical standard resin bead samples. The Safeguards Laboratory (NMCC) used their MAT 260 instrument and obtained the fractionation correction factor from resin bead standards provided with the TIGR-82 programme. Both PNC/TRP and NMCC reported problems with obtaining a sufficient ion beam intensity with the resin bead samples. This problem was overcome by both labs and further improvements in the loading and measurement techniques can be expected to yield even better results. It has been demonstrated that the resin bed sampling method can provide results of sufficient quality for safeguards purposes. 2 refs, 3 figs, 16 tabs

  1. Acid in perchloroethylene scrubber solutions used in HTGR fuel preparation processes. Analytical chemistry studies

    International Nuclear Information System (INIS)

    Lee, D.A.

    1979-02-01

    Acids and corrosion products in used perchloroethylene scrubber solutions collected from HTGR fuel preparation processes have been analyzed by several analytical methods to determine the source and possible remedy of the corrosion caused by these solutions. Hydrochloric acid was found to be concentrated on the carbon particles suspended in perchloroethylene. Filtration of carbon from the scrubber solutions removed the acid corrosion source in the process equipment. Corrosion products chemisorbed on the carbon particles were identified. Filtered perchloroethylene from used scrubber solutions contained practically no acid. It is recommended that carbon particles be separated from the scrubber solutions immediately after the scrubbing process to remove the source of acid and that an inhibitor be used to prevent the hydrolysis of perchloroethylene and the formation of acids

  2. Reclamation of zinc-contaminated soil using a dissolved organic carbon solution prepared using liquid fertilizer from food-waste composting

    International Nuclear Information System (INIS)

    Chiang, Po-Neng; Tong, Ou-Yang; Chiou, Chyow-San; Lin, Yu-An; Wang, Ming-Kuang; Liu, Cheng-Chung

    2016-01-01

    Highlights: • Nitrogen, phosphorus, and potassium contents in soil are substantially increased after the DOC washing. • The removal of Zn is dominated by proton replacement at pH 2.0, rather than by complexation with DOC. • The removal of Zn is dominated by DOC complexation between pH 3.0 and pH 5.0. - Abstract: A liquid fertilizer obtained through food-waste composting can be used for the preparation of a dissolved organic carbon (DOC) solution. In this study, we used the DOC solutions for the remediation of a Zn-contaminated soil (with Zn concentrations up to 992 and 757 mg kg −1 in topsoil and subsoil, respectively). We then determined the factors that affect Zn removal, such as pH, initial concentration of DOC solution, and washing frequency. Measurements using a Fourier Transform infrared spectrometer (FT-IR) revealed that carboxyl and amide were the major functional groups in the DOC solution obtained from the liquid fertilizer. Two soil washes using 1,500 mg L −1 DOC solution with a of pH 2.0 at 25 °C removed about 43% and 21% of the initial Zn from the topsoil and subsoil, respectively. Following this treatment, the pH of the soil declined from 5.4 to 4.1; organic matter content slightly increased from 6.2 to 6.5%; available ammonium (NH 4 + -N) content increased to 2.4 times the original level; and in the topsoil, the available phosphorus content and the exchangeable potassium content increased by 1.65 and 2.53 times their initial levels, respectively.

  3. Reclamation of zinc-contaminated soil using a dissolved organic carbon solution prepared using liquid fertilizer from food-waste composting

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Po-Neng [Experimental Forest, National Taiwan University, Chushan, Nantou County, 55750, Taiwan (China); Tong, Ou-Yang [Department of Environment Engineering, College of the Environment and Ecology, and The Key Laboratory of the Ministry of Education for Coastal and Wetland Ecosystem, Xiamen University, Xiamen (China); Chiou, Chyow-San; Lin, Yu-An [Department of Environmental Engineering, National Ilan University, Ilan 26047, Taiwan (China); Wang, Ming-Kuang [Department of Animal Science, National Ilan University, Ilan 26047, Taiwan (China); Liu, Cheng-Chung, E-mail: ccliu@niu.edu.tw [Department of Agricultural Chemistry, National Taiwan University, Taipei 10617, Taiwan (China)

    2016-01-15

    Highlights: • Nitrogen, phosphorus, and potassium contents in soil are substantially increased after the DOC washing. • The removal of Zn is dominated by proton replacement at pH 2.0, rather than by complexation with DOC. • The removal of Zn is dominated by DOC complexation between pH 3.0 and pH 5.0. - Abstract: A liquid fertilizer obtained through food-waste composting can be used for the preparation of a dissolved organic carbon (DOC) solution. In this study, we used the DOC solutions for the remediation of a Zn-contaminated soil (with Zn concentrations up to 992 and 757 mg kg{sup −1} in topsoil and subsoil, respectively). We then determined the factors that affect Zn removal, such as pH, initial concentration of DOC solution, and washing frequency. Measurements using a Fourier Transform infrared spectrometer (FT-IR) revealed that carboxyl and amide were the major functional groups in the DOC solution obtained from the liquid fertilizer. Two soil washes using 1,500 mg L{sup −1} DOC solution with a of pH 2.0 at 25 °C removed about 43% and 21% of the initial Zn from the topsoil and subsoil, respectively. Following this treatment, the pH of the soil declined from 5.4 to 4.1; organic matter content slightly increased from 6.2 to 6.5%; available ammonium (NH{sub 4}{sup +}-N) content increased to 2.4 times the original level; and in the topsoil, the available phosphorus content and the exchangeable potassium content increased by 1.65 and 2.53 times their initial levels, respectively.

  4. Numerical solution of the elastic non-axial contact between pellet and cladding of fuel rod in PWR

    International Nuclear Information System (INIS)

    Zymak, J.

    1987-08-01

    Elastic non-axial contacts between the pellet and the cladding of a fuel rod in a pressurized water reactor were calculated. The existence and the uniqueness of the solution were proved. The problem was approximated by the finite element method and quadratic programming was used for the solution. The results will be used in the solution of the probabilistic model of a fuel rod with non-axial pellets in a PWR. (author). 10 figs., 4 tabs., 10 refs

  5. Distributions of 14 elements on 63 absorbers from three simulant solutions (acid-dissolved sludge, acidified supernate, and alkaline supernate) for Hanford HLW Tank 102-SY

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1994-08-01

    As part of the Hanford Tank Waste Remediation System program at Los Alamos, we evaluated 63 commercially available or experimental absorber materials for their ability to remove hazardous components from high-level waste (HLW). These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. We tested these absorbers with three solutions prepared to simulate acid-dissolved sludge (pH 0.6), acidified supernate (pH 3.5), and alkaline supernate (pH 13.9) from underground storage tank 102-SY at the Hanford Reservation near Richland, Washington. To these simulants we added the appropriate radionuclides and used gamma spectrometry to measure fission products (Ce, Cs, Sr, Tc, and Y), actinides (U, Pu, and Am), and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr). For each of more than 2500 element/absorber/solution combinations, we measured distribution coefficients for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. Because we measured the sorption of many different elements, the tabulated results indicate those elements most likely to interfere with the sorption of elements of greater interest. On the basis of nearly 7500 measured distribution coefficients, we determined that many of these absorbers appear suitable for processing HLW. This study supersedes the previous version of LA-12654, in which results attributed to a solution identified as an alkaline supernate simulant were misleading because that solution contained insufficient hydroxide

  6. Uncertainty analysis of the nonideal competitive adsorption-donnan model: effects of dissolved organic matter variability on predicted metal speciation in soil solution.

    Science.gov (United States)

    Groenenberg, Jan E; Koopmans, Gerwin F; Comans, Rob N J

    2010-02-15

    Ion binding models such as the nonideal competitive adsorption-Donnan model (NICA-Donnan) and model VI successfully describe laboratory data of proton and metal binding to purified humic substances (HS). In this study model performance was tested in more complex natural systems. The speciation predicted with the NICA-Donnan model and the associated uncertainty were compared with independent measurements in soil solution extracts, including the free metal ion activity and fulvic (FA) and humic acid (HA) fractions of dissolved organic matter (DOM). Potentially important sources of uncertainty are the DOM composition and the variation in binding properties of HS. HS fractions of DOM in soil solution extracts varied between 14 and 63% and consisted mainly of FA. Moreover, binding parameters optimized for individual FA samples show substantial variation. Monte Carlo simulations show that uncertainties in predicted metal speciation, for metals with a high affinity for FA (Cu, Pb), are largely due to the natural variation in binding properties (i.e., the affinity) of FA. Predictions for metals with a lower affinity (Cd) are more prone to uncertainties in the fraction FA in DOM and the maximum site density (i.e., the capacity) of the FA. Based on these findings, suggestions are provided to reduce uncertainties in model predictions.

  7. Selecting enhancing solutions for electrokinetic remediation of dredged sediments polluted with fuel.

    Science.gov (United States)

    Rozas, F; Castellote, M

    2015-03-15

    In this paper a procedure for selecting the enhancing solutions in electrokinetic remediation experiments is proposed. For this purpose, dredged marine sediment was contaminated with fuel, and a total of 22 different experimental conditions were tested, analysing the influence of different enhancing solutions by using three commercial non-ionic surfactants, one bio-surfactant, one chelating agent, and one weak acid. Characterisation, microelectrophoretic and electrokinetic remediation trials were carried out. The results are explained on the basis of the interactions between the fuel, the enhancing electrolytes and the matrix. For one specific system, the electrophoretic zeta potential, (ζ), of the contaminated matrix in the solution was found to be related to the electroosmotic averaged ζ in the experiment and not to the efficiency in the extraction. This later was correlated to a parameter accounting for both contributions, the contaminant and the enhancing solution, calculated on the basis of differences in the electrophoretic ζ in different conditions which has allowed to propose a methodology for selection of enhancing solutions. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. Mesoporous yttria-zirconia and metal-yttria-zirconia solid solutions for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Mamak, M.; Coombs, N.; Ozin, G. [Toronto Univ., ON (Canada). Dept. of Chemistry

    2000-02-03

    A new class of binary mesoporous yttria-zirconia (YZ) and ternary mesoporous metal-YZ materials (M = electroactive Ni/Pt) is presented here that displays the highest surface area of any known form of yttria-stabilized zirconia. These mesoporous materials form as solid solutions and retain their structural integrity to 800 C, which bodes well for their possible utilization in fuel cells. (orig.)

  9. Used Fuel Logistics: Decades of Experience with transportation and Interim storage solutions

    Energy Technology Data Exchange (ETDEWEB)

    Orban, G.; Shelton, C.

    2015-07-01

    Used fuel inventories are growing worldwide. While some countries have opted for a closed cycle with recycling, numerous countries must expand their interim storage solutions as implementation of permanent repositories is taking more time than foreseen. In both cases transportation capabilities will have to be developed. AREVA TN has an unparalleled expertise with transportation of used fuel. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation model that was initially developed in the 1970s has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous “lessons learned” have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. AREVA TN has also an extensive experience with interim dry storage solutions in many countries on-site but also is working with partners to developed consolidated interim storage facility. Both expertise with storage and transportation contribute to safe, secure and smooth continuity of the operations. This paper will describe decades of experience with a very successful transportation program as well as interim storage solutions. (Author)

  10. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  11. Fuel effect on solution combustion synthesis of Co(Cr,Al)2O4 pigments

    International Nuclear Information System (INIS)

    Gilabert, J.; Palacios, M.D.; Sanz, V.; Mestre, S.

    2017-01-01

    The fuel effect on the synthesis of a ceramic pigment with a composition CoCr2−2ΨAl2ΨO4 (0≤Ψ≤1) by means of solution combustion synthesis process (SCS) has been studied. Three different fuels were selected to carry out the synthesis (urea, glycine and hexamethylentetramine (HMT)). Highly foamy pigments with very low density were obtained. Fd-3m spinel-type structure was obtained in all the experiments. Nevertheless, crystallinity and crystallite size of the spinels show significant differences with composition and fuel. The use of glycine along with the chromium-richest composition favours ion rearrangement to obtain the most ordered structure. Lattice parameter does not seem to be affected by fuel, although it evolves with Ψ according to Vegard's law. Colouring power in a transparent glaze shows important variations with composition. On the other hand, fuel effect presents a rather low influence since practically the same shades are obtained. However, it exerts certain effect on luminosity (L*). [es

  12. Method for dissolving ceramic beryllia

    International Nuclear Information System (INIS)

    Sands, A.E.

    1975-01-01

    A process is described for dissolving a nuclear fuel composition consisting of a sintered mass containing beryllia, a nuclear fuel selected from uranium and plutonium and a stabilizing agent, sintered at a temperature of at least 1500 0 C to a density of about 2.7 gs/cc. The process comprises contacting said sintered mass with a stoichiometric excess of lithium oxide dissolved or dispersed in a carrier selected from lithium hydroxide, sodium hydroxide or sodium nitrate at a temperature in the range 750--850 0 C to convert the beryllia to lithium beryllate and thereafter recovering the nuclear fuel content of said mass. (U.S.)

  13. Problems and solutions of the spent nuclear fuel (SNF) at Kozloduy NPP

    International Nuclear Information System (INIS)

    Jordanov, J.

    2003-01-01

    There are two options concerning spent nuclear fuel: to return it back to Russia for reprocessing or to store it on the site until we decide what to do with it. In both options prior to the shutting down of each reactor the Spent Fuel Pool thereto should be vacated (the filling in of the equipment at present is illustrated) and the Spent Fuel Storage Facility (SFSF) should also be vacated after the stop of the last nuclear facility on the site in order to be reequipped for permanent storage of the highly active wastes which will be returned in the country, if we submit the fuel for reprocessing; or of SNF, if we decide to leave them ultimately in Bulgaria. The difference is mainly in the quantities which will permanently remain here, respectively the volumes required for their storage and the funds necessary for the implementation of the processes. The pool volumes filling in both variants is also illustrated and the SFSF will be filled by 2008, if no fuel is transported.Costs of the SNF transport to Russia and investment costs of dry storage of SNF from pools 1 - 4 are present. The costs are visibly lower compared to those in the case of return of the fuel. However, these are only investments for construction and equipment of the buildings and storage containers. The costs related to their servicing are not included, and it should be taken into account that in approximately 50 years we will have to seek solution for their permanent storage. Despite the material costs to be incurred now for the implementation of the option with the return of the fuel, this is the more worthy way to resolve the problem. In accordance with the ethic principles in the nuclear energy, the burdens arising as a result of the use of nuclear facilities should be covered by the generation consuming the benefits from it

  14. Concentration-discharge relationships during an extreme event: Contrasting behavior of solutes and changes to chemical quality of dissolved organic material in the Boulder Creek Watershed during the September 2013 flood: SOLUTE FLUX IN A FLOOD EVENT

    Energy Technology Data Exchange (ETDEWEB)

    Rue, Garrett P. [Institute of Arctic and Alpine Research, University of Colorado, Boulder Colorado USA; Rock, Nathan D. [Institute of Arctic and Alpine Research, University of Colorado, Boulder Colorado USA; Gabor, Rachel S. [Institute of Arctic and Alpine Research, University of Colorado, Boulder Colorado USA; Pitlick, John [Department of Geography, University of Colorado, Boulder Colorado USA; Tfaily, Malak [Environmental Molecular Sciences Laboratory, Pacific Northwest National Laboratory, Richland Washington USA; McKnight, Diane M. [Institute of Arctic and Alpine Research, University of Colorado, Boulder Colorado USA

    2017-07-01

    During the week of September 10-17, 2013, close to 20 inches of rain fell across Boulder County, Colorado, USA. This rainfall represented a 1000-year event that caused massive hillslope erosion, landslides, and mobilization of sediments. The resultant stream flows corresponded to a 100-year flood. For the Boulder Creek Critical Zone Observatory (BC-CZO), this event provided an opportunity to study the effect of extreme rainfall on solute concentration-discharge relationships and biogeochemical catchment processes. We observed base cation and dissolved organic carbon (DOC) concentrations at two sites on Boulder Creek following the recession of peak flow. We also isolated three distinct fractions of dissolved organic matter (DOM) for chemical characterization. At the upper site, which represented the forested mountain catchment, the concentrations of the base cations Ca, Mg and Na were greatest at the peak flood and decreased only slightly, in contrast with DOC and K concentrations, which decreased substantially. At the lower site within urban corridor, all solutes decreased abruptly after the first week of flow recession, with base cation concentrations stabilizing while DOC and K continued to decrease. Additionally, we found significant spatiotemporal trends in the chemical quality of organic matter exported during the flood recession, as measured by fluorescence, 13C-NMR spectroscopy, and FTICR-MS. Similar to the effect of extreme rainfall events in driving landslides and mobilizing sediments, our findings suggest that such events mobilize solutes by the flushing of the deeper layers of the critical zone, and that this flushing regulates terrestrial-aquatic biogeochemical linkages during the flow recession.

  15. Rapid separation of pure 144Ce fraction from fuel dissolver solution for demonstration experiment on secular equilibrium

    International Nuclear Information System (INIS)

    Ashok Kumar, G.V.S.; Kumar, R.; Venkata Subramani, C.R.

    2015-01-01

    Radioactive equilibrium is a condition in which the activity ratio of parent to its daughter is maintained constant with time which occurs only when the parent half-life is greater than daughter half-life. It is transient equilibrium in the case of the ratio of their half-lives of parent to daughter being less than an order whereas it becomes secular equilibrium when it is more than an order. In the case of secular equilibrium, the ratio of the activities becomes unity whereas the same depends on the decay constants of the parent and daughter nuclide for the transient equilibrium. 144 Ce- 144 Pr pair is a good example for the demonstration of secular equilibrium

  16. Thermometric titration of a free acid and of uranyl in spent fuel element solutions

    International Nuclear Information System (INIS)

    Zamek, M.; Strafelda, F.

    1975-01-01

    A method was elaborated of determining nitric acid in the presence of uranyl nitrate in both aqueous and non-aqueous solutions using a pyridine aqueous solution as a titration agent, and of determining excess uranyl after a hydrogen peroxide addition by a further titration using the same agent. Even a hundred-fold excess of magnesium did not disturb the titration. The method is used in operating solution analyses in the extraction fuel reprocessing in the presence of a small amount of plutonium and of fission products. The reproducibility and accuracy of the method varied in the order of tens to units per cent depending on the concentration of components to be determined. The procedure is applicable for test volumes ranging between 0.1 and 10 ml in concentrations of 1 to 10 -3 M. (author)

  17. Supply Chain-Based Solution to Prevent Fuel Tax Evasion: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Capps, Gary J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Franzese, Oscar [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Lascurain, Mary Beth [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Siekmann, Adam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering and Transportation Sciences Division; Barker, Alan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Electrical and Electronics Systems Research Division

    2016-07-28

    The primary source of funding for the United States transportation system is derived from motor fuel and other highway use taxes. Loss of revenue attributed to fuel-tax evasion has been assessed to be somewhere between 1 billion and 3 billion per year. Any solution that addresses this problem needs to include not only the tax-collection agencies and auditors, but also the carriers transporting oil products and the carriers customers. This report presents a system developed by the Oak Ridge National Laboratory (ORNL) for the Federal Highway Administration which has the potential to reduce or eliminate many fuel-tax evasion schemes. The solution balances the needs of tax-auditors and those of the fuel-hauling companies and their customers. The system has three main components. The on-board subsystem combined sensors, tracking and communication devices, and software (the on-board Evidential Reasoning System, or obERS) to detect, monitor, and geo-locate the transfer of fuel among different locations. The back office sub-system (boERS) used self-learning algorithms to determine the legitimacy of the fuel loading and offloading (important for tax auditors) and detect potential illicit operations such as fuel theft (important for carriers and their customers, and may justify the deployment costs). The third sub-system, the Fuel Distribution and Auditing System or FDAS, is a centralized database, which together with a user interface allows tax auditors to query the data submitted by the fuel-hauling companies and correlate different parameters to quickly identify any anomalies. Industry partners included Barger Transport of Weber City, Virginia (fleet); Air-Weigh, of Eugene, Oregon (and their wires and harnesses); Liquid Bulk Tank (LBT) of Omaha, Nebraska (three five-compartment trailers); and Innovative Software Engineering (ISE) of Coralville, Iowa(on-board telematics device and back-office system). ORNL conducted a pilot test with the three instrumented vehicles

  18. Solutions obtained to international heat transfer benchmarking problems for nuclear fuel casks using Q/TRAN

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-02-01

    In 1985 Sandia National Laboratories participated in the Nuclear Energy Agency Committee on Reactor Physics (NEACRP) Specialists' Meeting on Heat Transfer Assessment of Transportation Packages. The objective of the meeting was to establish a set of model problems for use in comparing the performance of thermal analysis computer codes that may be used in the design of nuclear fuel shipping casks. The selected problems are to be used to compare code results for the thermal phenomena of conduction, convection, and radiation in cask-like problems. Two model problems were used in this study. The first problem required the determination of the steady-state temperatures of a 16 x 16 array of heated and unheated pins (representing fuel and control rod positions) of a simulated PWR fuel assembly. The second problem required the determination of transient temperatures of a finned surface (representing the external surface of a cask) subjected to an internal heat flux and to an external engulfing fire. Solutions to the problems were obtained with the code ''Q/TRAN.'' Solutions and descriptions of the necessary modeling techniques are given in this report

  19. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  20. Considerations in modelling the melting of fuel containing fission products and solute oxides

    International Nuclear Information System (INIS)

    Akbari, F.; Welland, M.J.; Lewis, B.J.; Thompson, W.T.

    2005-01-01

    It is well known that the oxidation of a defected fuel element by steam gives rise to an increase in O/U ratio with a consequent lowering of the incipient melting temperature. Concurrently, the hyperstoichiometry reduces the thermal conductivity thereby raising the centerline fuel pellet temperature for a fixed linear power. The development of fission products soluble in the UO 2 phase or, more important, the deliberate introduction of additive oxides in advanced CANDU fuel bundle designs further affects and generally lowers the incipient melting temperature. For these reasons, the modeling of the molten (hyperstoichiometric) UO 2 phase containing several solute oxides (ZrO 2 , Ln 2 O 3 and AnO 2 ) is advancing in the expectation of developing a moving boundary heat and mass transfer model aimed at better defining the limits of safe operating practice as burnup advances. The paper describes how the molten phase stability model is constructed. The redistribution of components across the solid-liquid interface that attends the onset of melting of a non-stoichiometric UO 2 containing several solutes will be discussed. The issues of how to introduce boundary conditions into heat transfer calculations consistent with the requirements of the Phase Rule will be addressed. The Stefan problem of a moving boundary associated with the solid/liquid interface sets this treatment apart from conventional heat and mass transfer problems. (author)

  1. Time series models for prediction the total and dissolved heavy metals concentration in road runoff and soil solution of roadside embankments

    Science.gov (United States)

    Aljoumani, Basem; Kluge, Björn; sanchez, Josep; Wessolek, Gerd

    2017-04-01

    Highways and main roads are potential sources of contamination for the surrounding environment. High traffic rates result in elevated heavy metal concentrations in road runoff, soil and water seepage, which has attracted much attention in the recent past. Prediction of heavy metals transfer near the roadside into deeper soil layers are very important to prevent the groundwater pollution. This study was carried out on data of a number of lysimeters which were installed along the A115 highway (Germany) with a mean daily traffic of 90.000 vehicles per day. Three polyethylene (PE) lysimeters were installed at the A115 highway. They have the following dimensions: length 150 cm, width 100 cm, height 60 cm. The lysimeters were filled with different soil materials, which were recently used for embankment construction in Germany. With the obtained data, we will develop a time series analysis model to predict total and dissolved metal concentration in road runoff and in soil solution of the roadside embankments. The time series consisted of monthly measurements of heavy metals and was transformed to a stationary situation. Subsequently, the transformed data will be used to conduct analyses in the time domain in order to obtain the parameters of a seasonal autoregressive integrated moving average (ARIMA) model. Four phase approaches for identifying and fitting ARIMA models will be used: identification, parameter estimation, diagnostic checking, and forecasting. An automatic selection criterion, such as the Akaike information criterion, will use to enhance this flexible approach to model building

  2. Inkjet-printed gold nanoparticle chemiresistors: Influence of film morphology and ionic strength on the detection of organics dissolved in aqueous solution

    International Nuclear Information System (INIS)

    Chow, Edith; Herrmann, Jan; Barton, Christopher S.; Raguse, Burkhard; Wieczorek, Lech

    2009-01-01

    The influence of film morphology on the performance of inkjet-printed gold nanoparticle chemiresistors has been investigated. Nanoparticles deposited from a single-solvent system resulted in a 'coffee ring'-like structure with most of the materials deposited at the edge. It was shown that the uniformity of the film could be improved if the nanoparticles were deposited from a mixture of solvents comprising N-methyl-2-pyrrolidone and water. Electrical conductivity measurements showed that both 'coffee ring' and 'flat' films were qualitatively similar suggesting that the films have similar nanoscale structures. To form the functional chemiresistor device, the 4-(dimethylamino)pyridine coating on the nanoparticle was exchanged with 1-hexanethiol to provide a hydrophobic sensing layer. The performance of 1-hexanethiol coated gold nanoparticle chemiresistors to small organic molecules, toluene, dichloromethane and ethanol dissolved in 1 M KCl in regard to changes in impedance and response times was unaffected by the film morphology. For larger hydrocarbons such as octane, the rate of uptake of the analyte into the film was significantly faster when the flatter nanoparticle film was used as opposed to the 'coffee ring' film which has a thicker edge. Furthermore, the presence of potassium and chloride ions in the solution media does not significantly affect the impedance of the nanoparticle film at 1 Hz (<2% variation in film impedance over more than four orders of magnitude change in ionic strength). However, the ionic strength of the media affected the partitioning of the analyte into the hydrophobic nanoparticle film. The response of the sensor was found to increase with an increased salt concentration due to a salting-out of the analyte from the solution

  3. Inkjet-printed gold nanoparticle chemiresistors: Influence of film morphology and ionic strength on the detection of organics dissolved in aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Chow, Edith [CSIRO Materials Science and Engineering, PO Box 218, Lindfield, NSW 2070 (Australia)], E-mail: Edith.Chow@csiro.au; Herrmann, Jan; Barton, Christopher S.; Raguse, Burkhard; Wieczorek, Lech [CSIRO Materials Science and Engineering, PO Box 218, Lindfield, NSW 2070 (Australia)

    2009-01-19

    The influence of film morphology on the performance of inkjet-printed gold nanoparticle chemiresistors has been investigated. Nanoparticles deposited from a single-solvent system resulted in a 'coffee ring'-like structure with most of the materials deposited at the edge. It was shown that the uniformity of the film could be improved if the nanoparticles were deposited from a mixture of solvents comprising N-methyl-2-pyrrolidone and water. Electrical conductivity measurements showed that both 'coffee ring' and 'flat' films were qualitatively similar suggesting that the films have similar nanoscale structures. To form the functional chemiresistor device, the 4-(dimethylamino)pyridine coating on the nanoparticle was exchanged with 1-hexanethiol to provide a hydrophobic sensing layer. The performance of 1-hexanethiol coated gold nanoparticle chemiresistors to small organic molecules, toluene, dichloromethane and ethanol dissolved in 1 M KCl in regard to changes in impedance and response times was unaffected by the film morphology. For larger hydrocarbons such as octane, the rate of uptake of the analyte into the film was significantly faster when the flatter nanoparticle film was used as opposed to the 'coffee ring' film which has a thicker edge. Furthermore, the presence of potassium and chloride ions in the solution media does not significantly affect the impedance of the nanoparticle film at 1 Hz (<2% variation in film impedance over more than four orders of magnitude change in ionic strength). However, the ionic strength of the media affected the partitioning of the analyte into the hydrophobic nanoparticle film. The response of the sensor was found to increase with an increased salt concentration due to a salting-out of the analyte from the solution.

  4. Separation of lanthanum from nuclear fuel solutions by high performance liquid chromatography

    International Nuclear Information System (INIS)

    Lazar, G. C.; Petre, M.; Androne, G.; Benga, A.

    2016-01-01

    This paper presents the separation of uranium, praseodymium and lanthanum from nuclear fuel solutions by high performance liquid chromatography (HPLC). The aim of this study is to establish a minimum concentration of lanthanum which can be analyzed by high performance liquid chromatography, and also to study the effect of uranium concentration on the separation of praseodymium and lanthanum. Optimum gradient mode was established for mixture standard stoc solutions with uranium in a concentration of 1 mg/ml, praseodymium and lanthanum in a concentration range of 1-5 μg/ml from each element. These conditions were applied for the separation of lanthanum from a nuclear fuel solution in which praseodymium and lanthanum were added in a concentration of 3 μg/ml from each element. The elution behavior of lanthanum as a function of the pH and the concentration of the mobile phase, using a mixture of 1-octanesulfonic acid sodium salt with a-hidroxyisobutiric acid is presented. (authors)

  5. Performance of direct alcohol fuel cells fed with mixed methanol/ethanol solutions

    Energy Technology Data Exchange (ETDEWEB)

    Wongyao, N. [The Joint Graduate School of Energy and Environment, King Mongkut' s University of Technology Thonburi, 126 Pracha-Uthit Rd., Bang Mod, Thung Khru, Bangkok 10140 (Thailand); Therdthianwong, A., E-mail: apichai.the@kmutt.ac.t [Fuel Cell and Hydrogen Research and Engineering Center, Clean Energy System Group, PDTI, King Mongkut' s University of Technology Thonburi, 126 Pracha-Uthit Rd., Bang Mod, Thung Khru, Bangkok 10140 (Thailand); Therdthianwong, S. [Department of Chemical Engineering, Faculty of Engineering, King Mongkut' s University of Technology Thonburi, 126 Pracha-Uthit Rd., Bang Mod, Thung Khru, Bangkok 10140 (Thailand)

    2011-07-15

    Research highlights: {yields} We examined the performance of direct alcohol fuel cells fed with mixed alcohol. {yields} PtRu-PtSn/C and PtRu/C as catalysts for mixed alcohol electrooxidation reaction. {yields} Misplace adsorption of ethanol on PtRu/C caused the cell performance drop. {yields} PtRu/C showed higher performance than PtRu-PtSn/C for mixed alcohol fuel. -- Abstract: In combining the advantages of both methanol and ethanol, direct alcohol fuel cells fed with mixed alcohol solutions (1 M methanol and 1 M ethanol in varying volume ratios) were tested for performance. Employing a PtRu-PtSn/C catalyst as anode, cell performance was found to diminish rapidly even at 2.5% by volume ethanol mixture. Further increase of ethanol exceeded 10%, the cell performance gradually decreased and finally approached that of direct ethanol fuel cells. The causes of the decrease in the cell performance were the slow electro-oxidation of ethanol and the misplaced adsorption of ethanol on PtRu/C. By comparing the PtRu-PtSn/C cell with the PtRu/C cell operated with mixed alcohol solutions, the cell using PtRu/C as an anode catalyst provided higher power density since more PtRu/C surface was available for methanol oxidation reaction and less ohmic resistance of PtRu/C than that of PtRu-PtSn/C. In order to reach optimization of DAFC performance fed with mixed alcohol, the electrocatalyst used for the anode must selectively adsorb an alcohol, especially ethanol.

  6. Performance of direct alcohol fuel cells fed with mixed methanol/ethanol solutions

    International Nuclear Information System (INIS)

    Wongyao, N.; Therdthianwong, A.; Therdthianwong, S.

    2011-01-01

    Research highlights: → We examined the performance of direct alcohol fuel cells fed with mixed alcohol. → PtRu-PtSn/C and PtRu/C as catalysts for mixed alcohol electrooxidation reaction. → Misplace adsorption of ethanol on PtRu/C caused the cell performance drop. → PtRu/C showed higher performance than PtRu-PtSn/C for mixed alcohol fuel. -- Abstract: In combining the advantages of both methanol and ethanol, direct alcohol fuel cells fed with mixed alcohol solutions (1 M methanol and 1 M ethanol in varying volume ratios) were tested for performance. Employing a PtRu-PtSn/C catalyst as anode, cell performance was found to diminish rapidly even at 2.5% by volume ethanol mixture. Further increase of ethanol exceeded 10%, the cell performance gradually decreased and finally approached that of direct ethanol fuel cells. The causes of the decrease in the cell performance were the slow electro-oxidation of ethanol and the misplaced adsorption of ethanol on PtRu/C. By comparing the PtRu-PtSn/C cell with the PtRu/C cell operated with mixed alcohol solutions, the cell using PtRu/C as an anode catalyst provided higher power density since more PtRu/C surface was available for methanol oxidation reaction and less ohmic resistance of PtRu/C than that of PtRu-PtSn/C. In order to reach optimization of DAFC performance fed with mixed alcohol, the electrocatalyst used for the anode must selectively adsorb an alcohol, especially ethanol.

  7. Dissolution studies of spent nuclear fuels

    International Nuclear Information System (INIS)

    1991-02-01

    To obtain quantitative data on the dissolution of high burnup spent nuclear fuel, dissolution study have been carried out at the Department of Chemistry, JAERI, from 1984 under the contract with STA entitled 'Reprocessing Test Study of High Burnup Fuel'. In this study PWR spent fuels of 8,400 to 36,100 MWd/t in averaged burnup were dissolved and the chemical composition and distribution of radioactive nuclides were measured for insoluble residue, cladding material (hull), off-gas and dissolved solution. With these analyses basic data concerning the dissolution and clarification process in the reprocessing plant were accumulated. (author)

  8. Interactions of hydrazine, ferrous sulfamate, sodium nitrite, and nitric acid in nuclear fuel processing solutions

    International Nuclear Information System (INIS)

    Gray, L.W.

    1977-03-01

    Hydrazine and ferrous sulfamate are used as reductants in a variety of nuclear fuel processing solutions. An oxidant, normally sodium nitrite, must frequently be added to these nitric acid solutions before additional processing can proceed. The interactions of these four chemicals have been studied under a wide variety of conditions using a 2/sup p/ factorial experimental design to determine relative reaction rates for desired reactions and side reactions. Evidence for a hydrazine-stabilized, sulfamic acid--nitrous acid intermediate was obtained; this intermediate can hydrolyze to ammonia or decompose to nitrogen. The oxidation of Fe 2+ by NO 2 - was shown to proceed at about the same rate as the scavenging of NO 2 - by sulfamic acid. Various side reactions are discussed

  9. Features of the Numerical Solution of Thermal Destruction Fuel Pins Problems in the Fast Reactor

    Science.gov (United States)

    Usov, E. V.; Butov, A. A.; Klimonov, I. A.; Chuhno, V. I.; Nikolaenko, A. V.; Zhdanov, V. S.; Pribaturin, N. A.; Strizhov, V. F.

    2017-11-01

    In this paper the description of the basic equations which can be used for calculation of melting of fuel and cladding of the fast reactor, moving of the melt on a fuel pin surface and its solidification is presented. The special attention is given speed of calculation algorithms and fidelity of the phenomena which are observed at a stage of severe accidents in fast reactors. For check of working capacity of initial models, numerical calculations of Stefan-type problems on front movement of melting/solidification in cylindrical geometry are presented. Comparison with the solutions received by known analytical methods is executed. For validation of the numerical realization of calculation algorithms the analysis is carried out and experiments in which melting of the model fuel pins of fast reactors was studied are chosen. On the basis of the chosen experiments calculation schemes taking into account initial and boundary conditions are prepared and modeling is performed. Modeling results are shown in the present paper. Estimation of calculation error of the basic physical parameters is done by results of the modeling and conclusions are drawn on a correctness of algorithms operation.

  10. Mixture of fuels for solution combustion synthesis of porous Fe3O4 powders

    Science.gov (United States)

    Parnianfar, H.; Masoudpanah, S. M.; Alamolhoda, S.; Fathi, H.

    2017-06-01

    The solution combustion synthesis of porous magnetite (Fe3O4) powders by a mixture of glycine and urea fuels was investigated concerning the thermodynamic aspects and powder characteristics. The adiabatic combustion temperature and combusted species were thermodynamically calculated as a function of the fuel to oxidant molar ratio (ϕ). The combustion behavior, phase evolution, porous structure and magnetic properties were characterized by thermal analysis, X-ray diffractometry, N2 adsorption-desorption, electron microscopy and vibrating sample magnetometry techniques. Nearly single phase Fe3O4 powders were synthesized by the mixture of fuels at ϕ values of 0.75 and 1. The as-combusted Fe3O4 powders at ϕ = 1 exhibited porous structure with the specific surface area of 83.4 m2/g. The highest saturation magnetization of 75.5 emu/g and the lowest coercivity of 84 Oe were achieved at ϕ = 1, due to the high purity and large crystallite size, inducing from the highest adiabatic combustion temperature.

  11. Safety of interim storage solutions of used nuclear fuel during extended term

    Energy Technology Data Exchange (ETDEWEB)

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

    2013-07-01

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

  12. Training Course of Experimental Chemistry in the Nuclear Fuel Cycle: Solid State and Solution Chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ju hyeong; Park, Kwangheon; Kim, Tae hoon; Park, Hyoung gyu; Kim, Jisu [Kyunghee University, Yongin (Korea, Republic of); Song, Hyuk jin [Dongguk University, Gyeongju (Korea, Republic of); Lee, Chan ki; Kang, Do kyu; Jeong, Hyeon jun [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    In this experimental study program in Tohoku University, basic experiments were done by the participants. First one is the hydrogen reduction experiment of the mixture of UO{sub 2} and ZrO{sub 2}. Second one is to observe microscopic structure of solid solution of UO{sub 2} and ZrO{sub 2} using SEM/EDX and XRD system, simulated fuel debris. Third one is milking process of {sup 239}Np from {sup 243}Am by solvent extraction using Tri-n-Octylamine (TOA). Last one is solvent extraction in PUREX by the simulated mixed aqueous solution of U, {sup 85}Sr and {sup 239}Np which is represented minor actinide elements included in the spent nuclear fuel. Uranium is separated from aqueous phase to organic phase during solvent extraction procedure using TBP and dodecane. Also, neptunium can be extracted to organic phase as nitric acid concentration change. The extraction behavior of neptunium is different by oxidation state in aqueous phase. The behavior of neptunium is represented as a combined form of these oxidation states in experiment. Therefore, because the oxidation states of neptunium can be controlled by controlling the concentration of nitric acid, the extractability of neptunium can be controlled.

  13. A method for recovering and separating palladium, technetium, rhodium and ruthenium contained in solutions resulting from nuclear fuel recycling

    International Nuclear Information System (INIS)

    Moore, R.H.

    1974-01-01

    The invention relates to a method for recovering and separating technetium and metals of the platinum group, i.e. palladium, rhodium and ruthenium existing as fission products. The method according to the invention is characterized by contacting a residuary acid aqueous solution provided by nuclear fuel recycling with successive carbon beds which have adsorbed different chelating agents specific for the metals to be recovered in order that said metals be selectively chelated and extracted from the solution. This method is suitable for recovering the above metals from solutions provided by reprocessing spent fuels [fr

  14. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

  15. Dissolution of intact UO2 pellet in batch and rotary dissolver conditions

    International Nuclear Information System (INIS)

    Jayendra Kumar Gelatar; Bijendra Kumar; Sampath, M.; Shekhar Kumar; Kamachi Mudali, U.; Natarajan, R.

    2015-01-01

    Comparative dissolution of intact un-irradiated UO 2 pellet of PHWR fuel dimensions was performed in batch and dynamic rotary dissolver conditions in aqueous nitric acid solutions at elevated temperatures. The extent of dissolution was estimated by determining the uranium concentration of the resulting aqueous solution. It was observed that rate of dissolution was much faster in dynamic conditions as compared to static batch conditions. (author)

  16. Dry storage technologies: Optimized solutions for spent fuels and vitrified residues

    International Nuclear Information System (INIS)

    Roland, Vincent; Verdier, Antoine; Sicard, Damien; Neider, Tara

    2006-01-01

    In many countries, fuel cycle and waste policies influence the way operators organize waste management. These policies help drive progress and improvements in areas such as waste minimization programs, conditioning or industrial transformation before final or intermediate conditioning. The criteria that lead to different choices include economic factors, the presence or absence of a wide range of options such as transport, and reprocessing and recycling policies. The current international trend towards expanding Spent Fuel Interim Dry Storage capabilities goes with an improvement of the performance of the proposed systems which have to accommodate Spent fuel Assemblies characterized by ever increasing burn-up, fissile isotopes contents, thermal releases, and total inventory. Due to heterogeneous worldwide reactor pools and specific local constraints the proposed solutions have also to cope with a wide variety of fuel design. The Spent Fuel Assemblies stored temporarily for cooling may have to be transported either to reprocessing facilities or to interim storage facilities before direct disposal; it is the reason why the retrievability, including or not the need of transportation of the proposed systems, is often specified by the utilities for the design of their storage systems and sometimes required by law. In most cases, the producers of spent fuel require a large capacity that is cumulated over many years, each reload at a time. Then the key criterion is maximum modularity. Furthermore, the up front capital costs required for this type of solution has to be attractive for the investor. Two solutions, dual purpose metal casks of the TN TM 24 family or dual purpose or single purpose concrete shielded welded canisters such as NUHOMS R , implemented by COGEMA LOGISTICS, and TRANSNUCLEAR Inc. offer flexibility and modularity and have been adapted also to quite different fuels. Among what influences the choice, we can consider: - In favor of metal casks: Minimal

  17. Photocatalytic degradation of H2S aqueous media using sulfide nanostructured solid-solution solar-energy-materials to produce hydrogen fuel.

    Science.gov (United States)

    Lashgari, Mohsen; Ghanimati, Majid

    2018-03-05

    H 2 S is a corrosive, flammable and noxious gas, which can be neutralized by dissolving in alkaline media and employed as H 2 -source by utilizing inside semiconductor-assisted/photochemical reactors. Herein, through a facile hydrothermal route, a ternary nanostructured solid-solution of iron, zinc and sulfur was synthesized in the absence and presence of Ag-dopant, and applied as efficient photocatalyst of hydrogen fuel production from H 2 S media. The effect of pH on the photocatalyst performance was scrutinized and the maximum activity was attained at pH=11, where HS - concentration is high. BET, diffuse reflectance and photoluminescence studies indicated that the ternary solid-solution photocatalyst, in comparison to its solid-solvent (ZnS), has a greater surface area, stronger photon absorption and less charge recombination, which justify its superiority. Moreover, the effect of silver-dopant on the photocatalyst performance was examined. The investigations revealed that although silver could boost the absorption of photons and increase the surface area, it could not appreciably enhance the photocatalyst performance due to its weak influence on retarding the charge-recombination process. Finally, the phenomenon was discussed in detail from mechanistic viewpoint. Copyright © 2017 Elsevier B.V. All rights reserved.

  18. Mass transfer in fuel cells. [electron microscopy of components, thermal decomposition of Teflon, water transport, and surface tension of KOH solutions

    Science.gov (United States)

    Walker, R. D., Jr.

    1973-01-01

    Results of experiments on electron microscopy of fuel cell components, thermal decomposition of Teflon by thermogravimetry, surface area and pore size distribution measurements, water transport in fuel cells, and surface tension of KOH solutions are described.

  19. ICPP custom dissolver explosion recovery

    International Nuclear Information System (INIS)

    Demmer, R.; Hawk, R.

    1992-01-01

    This paper discusses the recovery from the February 9, 1991, small scale explosion in a custom processing dissolver at the Idaho Chemical Processing Plant (ICPP) a Department of Energy facility at the Idaho National Engineering Laboratory. The custom processing facility is a limited production area designed to recover unirradiated uranium fuel. A small amount of the nuclear material received and stored at the ICPP is unique and incompatible with the major head end dissolution processes. Custom processing is a small scale dissolution facility for processing these materials in an economical fashion in the CPP-627 hot chemistry laboratory. Two glass dissolvers were contained in a large walk in hood area. Utilities for dissolution and connections to the major ICPP uranium separation facility were provided. The fuel processing operations during this campaign involved dissolving uranium metal, uranium oxides, and uranium/fissium alloy in nitric acid

  20. Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

    International Nuclear Information System (INIS)

    St John, C.M.

    1977-04-01

    An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

  1. Dissolved gases

    International Nuclear Information System (INIS)

    Heaton, T.H.E.

    1981-01-01

    The concentrations of gaseous nitrogen, argon, oxygen and helium dissolved in groundwater are often different from their concentrations in rain and surface waters. These differences reflect changes in the gas content occurring after rain or surface water, having infiltrated into the ground, become isolated from equilibrium contact with the atmosphere. A study of these changes can give insight into the origin and subsequent subsurface history of groundwater. Nitrogen and argon concentrations for many groundwaters in southern Africa indicate that excess air is added to water during infiltration. The amount of excess air is believed to reflect the physical structure of the unsaturated zone and the climate of the recharge area. Since nitrogen and argon are essentially conservative in many aquifer environments in South Africa, their concentrations can be used in distinguishing grondwaters of different recharge origins. In some areas the high helium content of the groundwater suggests that much of the helium is derived through migration from a source outside (e.g. below) the aquifer itself. Radiogenic helium concentrations nevertheless show, in two artesian aquifers, a close linear relationship to the radiocarbon age of the groundwater. This indicates a uniformity in the factors responsible for the accumulation of helium, and suggests that in these circumstances helium data can be used to give information on the age of very old groundwater. In some groundwater dissolved oxygen concentrations are found to decrease with increasing groundwater age. Whilst the rate of decrease may be very different for different aquifers, the field measurement of oxygen may be useful in preliminary surveys directed toward the location of recharge areas

  2. Study of influence of fuel on dielectric and ferroelectric properties of bismuth titanate ceramics synthesized using solution based combustion technique

    International Nuclear Information System (INIS)

    Subohi, Oroosa; Malik, M M; Kurchania, Rajnish; Kumar, G S

    2015-01-01

    The effect of fuel characteristics on the processing and properties of bismuth titanate (BIT) ceramics obtained by solution combustion route using different fuels are reported in this paper. Dextrose, urea and glycine were used as fuel in this study. The obtained bismuth titanate ceramics were characterized by using XRD, SEM at different stages of sample preparation. It was observed that BIT obtained by using dextrose as fuel shows higher dielectric constant and higher remnant polarization due to smaller grain size and lesser c-axis growth as compared to the samples with urea and glycine as fuel. The electrical behavior of the samples with respect to temperature and frequency was also investigated to understand relaxation phenomenon. (paper)

  3. Adsorption of mercury from aqueous solutions using palm oil fuel ash as an adsorbent - batch studies

    Science.gov (United States)

    Imla Syafiqah, M. S.; Yussof, H. W.

    2018-03-01

    Palm oil fuel ash (POFA) is one of the most abundantly produced waste materials. POFA is widely used by the oil palm industry which was collected as ash from the burning of empty fruit bunches fiber (EFB) and palm oil kernel shells (POKS) in the boiler as fuel to generate electricity. Mercury adsorption was conducted in a batch process to study the effects of contact time, initial Hg(II) ion concentration, and temperature. In this study, POFA was prepared and used for the removal of mercury(II) ion from the aqueous phase. The effects of various parameters such as contact time (0- 360 min), temperature (15 – 45 °C) and initial Hg(II) ion concentration (1 – 5 mg/L) for the removal of Hg(II) ion were studied in a batch process. The surface characterization was examined by scanning electron microscopy (SEM) and particle size distribution analysis. From this study, it was found that the highest Hg(II) ion removal was 99.60 % at pH 7, contact time of 4 h, initial Hg(II) ion concentration of 1 mg/L, adsorbent dosage 0.25 g and agitation speed of 100 rpm. The results implied that POFA has the potential as a low-cost and environmental friendly adsorbent for the removal of mercury from aqueous solution.

  4. Freely dissolved concentrations of anionic surfactants in seawater solutions: optimization of the non-depletive solid-phase microextraction method and application to linear alkylbenzene sulfonates.

    NARCIS (Netherlands)

    Rico Rico, A.; Droge, S.T.J.; Widmer, D.; Hermens, J.L.M.

    2009-01-01

    A solid-phase microextraction method (SPME) has been optimized for the analysis of freely dissolved anionic surfactants, namely linear alkylbenzene sulfonates (LAS), in seawater. An effect of the thermal conditioning treatment on the polyacrylate fiber coating was demonstrated for both uptake

  5. Modeling of the anode side of a direct methanol fuel cell with analytical solutions

    International Nuclear Information System (INIS)

    Mosquera, Martin A.; Lizcano-Valbuena, William H.

    2009-01-01

    In this work, analytical solutions were derived (for any methanol oxidation reaction order) for the profiles of methanol concentration and proton current density, by assuming diffusion mass transport mechanism, Tafel kinetics, and fast proton transport in the anodic catalyst layer of a direct methanol fuel cell. An expression for the Thiele modulus that allows to express the anodic overpotential as a function of the cell current and kinetic and mass transfer parameters was obtained. For high cell current densities, it was found that the Thiele modulus (φ 2 ) varies quadratically with cell current density; yielding a simple correlation between anodic overpotential and cell current density. Analytical solutions were derived for the profiles of both local methanol concentration in the catalyst layer and local anodic current density in the catalyst layer. Under the assumptions of the model presented here, in general, the local methanol concentration in the catalyst layer cannot be expressed as an explicit function of the position in the layer. In spite of this, the equations presented here for the anodic overpotential allow the derivation of new semi-empirical equations

  6. Convincing about the advanced use of nuclear energy closing the fuel cycle: from a burden to a solution

    International Nuclear Information System (INIS)

    Neau, Henry Jacques

    2007-01-01

    France has associated a closed fuel cycle with its nuclear program, and developed the corresponding treatment recycling capabilities accordingly. This choice was recently consolidated by law. according to the sustainable management of radioactive materials and waste act of June 2006, the volume and radio toxicity reduction of nuclear waste is an objective that can notably be reached with their treatment and conditioning. Presently, used fuel valuable components (U and Pu) are recycled into MOX fuel and RepU, when fission products are conditioned under an extremely solid and resistant form which cannot disperse and dissolve in the environment (High Level Vitrified Waste). Safety and waste minimisation remain the AREVA constant objective. Presently operated treatment and recycling AREVA NC facilities are using mature industrial technologies, which address environment preservation and non proliferation concerns. This french national choice requires a permanent global acceptance strategy towards politicians, media, associations and more generally public opinion: to. be accepted, in needs to be understood. Transparency, dialogue and information are keywords for AREVA NC to be sure that closing the fuel cycle is considered as the best option available now for responsibly managing the waste, respecting the environment, preserving the resource and securing the future. Partnering in this Global Acceptance policy with other countries and customers, who already rely- or plan to do so - on this recycling strategy is both a reality and a permanent axis of development for AREVA NC

  7. Agro-fuels and climate. Why is it not a good solution?

    International Nuclear Information System (INIS)

    2015-10-01

    This publication explains why agro-fuels raise various problems: deforestation due to a change of land use, impact on climate due to deforestation and higher CO 2 emission by colza-based fuel than by conventional diesel fuel, food safety due to the evolution of land uses which result in higher food prices and land grabbing. It also outlines that agro-fuels are a burden for public finances. Possibilities offered by agro-fuels of third generation are evoked

  8. Solution combustion synthesis of strontium aluminate, SrAl2O4, powders: single-fuel versus fuel-mixture approach.

    Science.gov (United States)

    Ianoş, Robert; Istratie, Roxana; Păcurariu, Cornelia; Lazău, Radu

    2016-01-14

    The solution combustion synthesis of strontium aluminate, SrAl2O4, via the classic single-fuel approach and the modern fuel-mixture approach was investigated in relation to the synthesis conditions, powder properties and thermodynamic aspects. The single-fuel approach (urea or glycine) did not yield SrAl2O4 directly from the combustion reaction. The absence of SrAl2O4 was explained by the low amount of energy released during the combustion process, in spite of the highly negative values of the standard enthalpy of reaction and standard Gibbs free energy. In the case of single-fuel recipes, the maximum combustion temperatures measured by thermal imaging (482 °C - urea, 941 °C - glycine) were much lower than the calculated adiabatic temperatures (1864 °C - urea, 2147 °C - glycine). The fuel-mixture approach (urea and glycine) clearly represented a better option, since (α,β)-SrAl2O4 resulted directly from the combustion reaction. The maximum combustion temperature measured in the case of a urea and glycine fuel mixture was the highest one (1559 °C), which was relatively close to the calculated adiabatic temperature (1930 °C). The addition of a small amount of flux, such as H3BO3, enabled the formation of pure α-SrAl2O4 directly from the combustion reaction.

  9. ICPP custom dissolver explosion recovery

    International Nuclear Information System (INIS)

    Demmer, R.; Hawk, R.

    1992-01-01

    This report discusses the recovery from the February 9, 1991 small scale explosion in a custom processing dissolver at the Idaho Chemical Processing Plant. Custom processing is a small scale dissolution facility which processes nuclear material in an economical fashion. The material dissolved in this facility was uranium metal, uranium oxides, and uranium/fissium alloy in nitric acid. The paper explained the release of fission material, and the decontamination and recovery of the fuel material. The safety and protection procedures were also discussed. Also described was the chemical analysis which was used to speculate the most probable cause of the explosion. (MB)

  10. FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

    1976-08-01

    A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

  11. Development of a real-time fuel cell stack modelling solution with integrated test rig interface for the generic fuel cell modelling environment (GenFC) software

    Energy Technology Data Exchange (ETDEWEB)

    Fraser, S.D.; Monsberger, M.; Hacker, V. [Graz Univ. of Technology, Graz (Austria). Christian Doppler Laboratory for Fuel Cell Systems; Gubner, A.; Reimer, U. [Forschungszentrum Julich, Julich (Germany)

    2007-07-01

    Since the late 1980s, numerous FC models have been developed by scientists and engineers worldwide to design, control and optimize fuel cells (FCs) and fuel cell (FC) power systems. However, state-of-the-art FC models have only a small range of applications within the versatile field of FC modelling. As fuel cell technology approaches commercialization, the scientific community is faced with the challenge of providing robust fuel cell models that are compatible with established processes in industrial product development. One such process, known as Hardware in the Loop (HiL), requires real-time modelling capability. HiL is used for developing and testing hardware components by adding the complexity of the related dynamic systems with mathematical representations. Sensors and actuators are used to interface simulated and actual hardware components. As such, real-time fuel cell models are among the key elements in the development of the Generic Fuel Cell Modelling Environment (GenFC) software. Six European partners are developing GenFC under the support of the Sixth European Framework Programme for Research and Technological Development (FP6). GenFC is meant to increase the use of fuel cell modelling for systems design and to enable cost- and time-efficient virtual experiments for optimizing operating parameters. This paper presented an overview of the GenFC software and the GenFC HiL functionality. It was concluded that GenFC is going to be an extendable software tool providing FC modelling techniques and solutions to a wide range of different FC modelling applications. By combining the flexibility of the GenFC software with this HiL-specific functionality, GenFC is going to promote the use of FC model-based HiL technology in FC system development. 9 figs.

  12. recovery of enriched uranium from waste solution obtained from fuel fabrication laboratories

    International Nuclear Information System (INIS)

    Othman, S.H.A.

    2003-01-01

    reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of uranium (19.7% enriched with 235 U) from highly impure solution . the processing of uranium compounds for atomic energy project especially in FMPP(Egyptian fuel manufacture pilot plant) gives rise to a variety of wastes in which the uranium content is of considerable importance. the recovery of uranium from concentrated mother liquors produced from ADU (ammonium diuranate ) precipitation, as well as those due to ADU washing is studied in this work. column of poly-trifluoro-monochloro-ethilene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains uranium .impurities are eluted with 6.5 M HCl, and the uranium is eluted with water and the recovery of uranium is better than 94%. A mathematical model was suggested to stimulate the sorption process of uranium ions (or any other ion ) by column of solvent impregnated resin containing organic extractant (the same as the previous column) . An excellent agreement was founded between the experimental results and the mathematical model

  13. Proton exchange membrane fuel cell for cooperating households: A convenient combined heat and power solution for residential applications

    International Nuclear Information System (INIS)

    Cappa, Francesco; Facci, Andrea Luigi; Ubertini, Stefano

    2015-01-01

    In this paper we compare the technical and economical performances of a high temperature proton exchange membrane fuel cell with those of an internal combustion engine for a 10 kW combined heat and power residential application. In a view of social innovation, this solution will create new partnerships of cooperating families aiming to reduce the energy consumption and costs. The energy system is simulated through a lumped model. We compare, in the Italian context, the total daily operating cost and energy savings of each system with respect to the separate purchase of electricity from the grid and production of the thermal energy through a standard boiler. The analysis is carried out with the energy systems operating with both the standard thermal tracking and an optimized management. The latter is retrieved through an optimization methodology based on the graph theory. We show that the internal combustion engine is much more affected by the choice of the operating strategy with respect to the fuel cell, in terms long term profitability. Then we conduct a net present value analysis with the aim of evidencing the convenience of using a high temperature proton exchange membrane fuel cell for cogeneration in residential applications. - Highlights: • Fuel cells are a feasible and economically convenient solution for residential CHP. • Control strategy is fundamental for the economical performance of a residential CHP. • Flexibility is a major strength of the fuel cell CHP.

  14. Fuel Continuous Mixer ? an Approach Solution to Use Straight Vegetable Oil for Marine Diesel Engines

    Directory of Open Access Journals (Sweden)

    Đặng Van Uy

    2018-03-01

    Full Text Available The vegetable oil is well known as green fuel for diesel engines due to its low sunphur content and renewable stock. However, there are some problems raising when vegetable oil is used as fuel for diesel engines such as highly effected by cold weather, lower general efficiency, separation in layer if mixed with diesel oil and so on. To overcome that disadvantiges, the authors propose a new idea that to use a continuous fuel mixer to blend vegetable oil with diesel oil to make so called a mixed fuel supplying to diesel engines inline. In order to ensure a quality of the mixed fuel created by continuous mixer, a homogeneous testing was introduced with believable results. Then, the continuous mixer has been installed into fuel supply system of diesel engine 6LU32 at a lab of Vietnam Maritime University in terms of checking a real operation of the fuel continuous mixer with diesel engine.

  15. Determination of U and Pu in highly radioactive solutions. Application to the analysis of fuel reprocessing solutions

    International Nuclear Information System (INIS)

    Denis, Alphonse

    Various devices intended for the X-ray fluorescence analysis of U and Pu in radioactive solutions were developed in a hot cell: a device for sample deposition and preparation; a measurement unit. The problems investigated are described [fr

  16. Supply Chain Based Solution to Prevent Fuel Tax Evasion: Proof of Concept Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Capps, Gary J [ORNL; Lascurain, Mary Beth [ORNL; Franzese, Oscar [ORNL; Earl, Dennis Duncan [ORNL; West, David L [ORNL; McIntyre, Timothy J [ORNL; Chin, Shih-Miao [ORNL; Hwang, Ho-Ling [ORNL; Connatser, Raynella M [ORNL; Lewis Sr, Samuel Arthur [ORNL; Moore, Sheila A [ORNL

    2011-12-01

    The goal of this research was to provide a proof-of-concept (POC) system for preventing non-taxable (non-highway diesel use) or low-taxable (jet fuel) petrochemical products from being blended with taxable fuel products and preventing taxable fuel products from cross-jurisdiction evasion. The research worked to fill the need to validate the legitimacy of individual loads, offloads, and movements by integrating and validating, on a near-real-time basis, information from global positioning system (GPS), valve sensors, level sensors, and fuel-marker sensors.

  17. Removal of copper from aqueous solution by electrodeposition in cathode chamber of microbial fuel cell.

    Science.gov (United States)

    Tao, Hu-Chun; Liang, Min; Li, Wei; Zhang, Li-Juan; Ni, Jin-Ren; Wu, Wei-Min

    2011-05-15

    Based on energetic analysis, a novel approach for copper electrodeposition via cathodic reduction in microbial fuel cells (MFCs) was proposed for the removal of copper and recovery of copper solids as metal copper and/or Cu(2)O in a cathode with simultaneous electricity generation with organic matter. This was examined by using dual-chamber MFCs (chamber volume, 1L) with different concentrations of CuSO(4) solution (50.3 ± 5.8, 183.3 ± 0.4, 482.4 ± 9.6, 1007.9 ± 52.0 and 6412.5 ± 26.7 mg Cu(2+)/L) as catholyte at pH 4.7, and different resistors (0, 15, 390 and 1000 Ω) as external load. With glucose as a substrate and anaerobic sludge as an inoculum, the maximum power density generated was 339 mW/m(3) at an initial 6412.5 ± 26.7 mg Cu(2+)/L concentration. High Cu(2+) removal efficiency (>99%) and final Cu(2+) concentration below the USA EPA maximum contaminant level (MCL) for drinking water (1.3mg/L) was observed at an initial 196.2 ± 0.4 mg Cu(2+)/L concentration with an external resistor of 15 Ω, or without an external resistor. X-ray diffraction analysis confirmed that Cu(2+) was reduced to cuprous oxide (Cu(2)O) and metal copper (Cu) on the cathodes. Non-reduced brochantite precipitates were observed as major copper precipitates in the MFC with a high initial Cu(2+) concentration (0.1M) but not in the others. The sustainability of high Cu(2+) removal (>96%) by MFC was further examined by fed-batch mode for eight cycles. Copyright © 2011 Elsevier B.V. All rights reserved.

  18. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  19. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; McGinley, J.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements P.O.Box 2340, Karlsruhe, D-76125 (Germany)

    2008-07-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  20. FY12 Final Report for PL10-Mod Separations-PD12: Electrochemically Modulated Separation of Plutonium from Dilute and Concentrated Dissolver Solutions for Analysis by Gamma Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Pratt, Sandra H.; Arrigo, Leah M.; Duckworth, Douglas C.; Cloutier, Janet M.; Breshears, Andrew T.; Schwantes, Jon M.

    2013-05-01

    Accurate and timely analysis of plutonium in spent nuclear fuel is critical in nuclear safeguards for detection of both protracted and rapid plutonium diversions. Gamma spectroscopy is a viable method for accurate and timely measurements of plutonium provided that the plutonium is well separated from the interfering fission and activation products present in spent nuclear fuel. Electrochemically modulated separation (EMS) is a method that has been used successfully to isolate picogram amounts of Pu from nitric acid matrices. With EMS, Pu adsorption may be turned “on” and “off” depending on the applied voltage, allowing for collection and stripping of Pu without the addition of chemical reagents. In this work, we have scaled up the EMS process to isolate microgram quantities of Pu from matrices encountered in spent nuclear fuel during reprocessing. Several challenges have been addressed including surface area limitations, radiolysis effects, electrochemical cell performance stability, and chemical interferences. After these challenges were resolved, 6 µg Pu was deposited in the electrochemical cell with approximately an 800-fold reduction of fission and activation product levels from a spent nuclear fuel sample. Modeling showed that these levels of Pu collection and interference reduction may not be sufficient for Pu detection by gamma spectroscopy. The main remaining challenges are to achieve a more complete Pu isolation and to deposit larger quantities of Pu for successful gamma analysis of Pu. If gamma analyses of Pu are successful, EMS will allow for accurate and timely on-site analysis for enhanced Pu safeguards.

  1. Improving the electrocatalytic properties of Pd-based catalyst for direct alcohol fuel cells: effect of solid solution.

    Science.gov (United States)

    Wen, Cuilian; Wei, Ying; Tang, Dian; Sa, Baisheng; Zhang, Teng; Chen, Changxin

    2017-07-07

    The tolerance of the electrode against the CO species absorbed upon the surface presents the biggest dilemma of the alcohol fuel cells. Here we report for the first time that the inclusion of (Zr, Ce)O 2 solid solution as the supporting material can significantly improve the anti-CO-poisoning as well as the activity of Pd/C catalyst for ethylene glycol electro-oxidation in KOH medium. In particular, the physical origin of the improved electrocatalytic properties has been unraveled by first principle calculations. The 3D stereoscopic Pd cluster on the surface of (Zr, Ce)O 2 solid solution leads to weaker Pd-C bonding and smaller CO desorption driving force. These results support that the Pd/ZrO 2 -CeO 2 /C composite catalyst could be used as a promising effective candidate for direct alcohol fuel cells application.

  2. Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Jegou, C., E-mail: christophe.jegou@cea.f [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Caraballo, R.; De Bonfils, J.; Broudic, V.; Peuget, S. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Vercouter, T. [Commissariat a l' Energie Atomique (CEA), Saclay Reasearch Center, B.P. 11, F-91191 Gif-sur-Yvette Cedex (France); Roudil, D. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France)

    2010-04-01

    The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2 (registered) ) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 deg. C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO{sub 2} matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO{sub 2} grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH){sub 4(am)} phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.

  3. Approximate solutions of pulse transport in turbulent flow in narrow fuel element bundle geometries, using the FE method

    International Nuclear Information System (INIS)

    Kaiser, H.G.

    1985-01-01

    The author is concerned with the flow conditions in case of narrow fuel element grids of pressurised-water reactors. Starting from the mathematical formulation of the flow processes for incompressible, isothermal flows, models of the turbulence characteristics are being developed. Besides turbulence models, and network structure the finite element method is treated as numeric solution process. Finally the results are summarized and discussed. (HAG) [de

  4. Analytical solution and experimental validation of the energy management problem for fuel cell hybrid vehicles

    NARCIS (Netherlands)

    P.P.J. van den Bosch; Edwin Tazelaar; M. Grimminck; Stijn Hoppenbrouwers; Bram Veenhuizen

    2011-01-01

    The objective of an energy management strategy for fuel cell hybrid propulsion systems is to minimize the fuel needed to provide the required power demand. This minimization is defined as an optimization problem. Methods such as dynamic programming numerically solve this optimization problem.

  5. Analytical solution of the energy management for fuel cell hybrid propulsion systems

    NARCIS (Netherlands)

    P.P.J. van den Bosch; E. Tazelaar; Bram Veenhuizen

    2012-01-01

    The objective of an energy management strategy for fuel cell hybrid propulsion systems is to minimize the fuel needed to provide the required power demand. This minimization is defined as an optimization problem. Methods such as dynamic programming numerically solve this optimization problem.

  6. Closed-form solution of a two-dimensional fuel temperature model for TRIGA-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rivard, J B [Sandia Laboratories (United States)

    1974-07-01

    If azimuthal power density variations are ignored, the steady-state temperature distribution within a TRIGA-type fuel element is given by the solution of the Poisson equation in two dimensions (r and z) . This paper presents a closed-form solution of this equation as a function of the axial and radial power density profiles, the conductivity of the U-ZrH, the inlet temperature, specific heat and flow rate of the coolant, and the overall heat transfer coefficient. The method begins with the development of a system of linear ordinary differential equations describing mass and energy balances in the fuel and coolant. From the solution of this system, an expression for the second derivative of the fuel temperature distribution in the axial (z) direction is found. Substitution of this expression into the Poisson equation for T(r,z) reduces it from a partial differential equation to an ordinary differential equation in r, which is subsequently solved in closed-form. The results of typical calculations using the model are presented. (author)

  7. Stress Analysis of a TRISO Coated Particle Fuel by Using ABAQUS Finite Element Visco-Elastoplastic Solutions

    International Nuclear Information System (INIS)

    Cho, Moon Sung; Kim, Y. M.; Lee, Y. W.

    2006-01-01

    The fundamental design for a gas-cooled reactor relies on an understanding of the behavior of a coated particle fuel. KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) Project since 2004, is developing a fuel performance analysis code for a VHTR named COPA (COated Particle fuel Analysis). A validation of COPA was attempted by comparing its benchmark results with the visco-elastic solutions obtained from the ABAQUS code calculations for the IAEA-CRP-6 TRISO coated particle benchmark problems involving a creep, swelling, and pressure. However, the ABAQUS finite element model used for the above-mentioned analysis did not consider the material nonlinearity of the SiC coating layer that showed stress levels higher than the assumed yield point of the material. In this study, a consideration of the material nonlinearity is included in the ABAQUS model to obtain the visco-elastoplastic solutions and the results are compared with the visco-elastic solutions obtained from the previous ABAQUS model

  8. Plant induced changes in concentrations of caesium, strontium and uranium in soil solution with reference to major ions and dissolved organic matter

    International Nuclear Information System (INIS)

    Takeda, Akira; Tsukada, Hirofumi; Takaku, Yuichi; Akata, Naofumi; Hisamatsu, Shun'ichi

    2008-01-01

    For a better understanding of the soil-to-plant transfer of radionuclides, their behavior in the soil solution should be elucidated, especially at the interface between plant roots and soil particles, where conditions differ greatly from the bulk soil because of plant activity. This study determined the concentration of stable Cs and Sr, and U in the soil solution, under plant growing conditions. The leafy vegetable komatsuna (Brassica rapa L.) was cultivated for 26 days in pots, where the rhizosphere soil was separated from the non-rhizosphere soil by a nylon net screen. The concentrations of Cs and Sr in the rhizosphere soil solution decreased with time, and were controlled by K + NH 4 + and Ca, respectively. On the other hand, the concentration of U in the rhizosphere soil solution increased with time, and was related to the changes of DOC; however, this relationship was different between the rhizosphere and non-rhizosphere soil

  9. Plant induced changes in concentrations of caesium, strontium and uranium in soil solution with reference to major ions and dissolved organic matter.

    Science.gov (United States)

    Takeda, Akira; Tsukada, Hirofumi; Takaku, Yuichi; Akata, Naofumi; Hisamatsu, Shun'ichi

    2008-06-01

    For a better understanding of the soil-to-plant transfer of radionuclides, their behavior in the soil solution should be elucidated, especially at the interface between plant roots and soil particles, where conditions differ greatly from the bulk soil because of plant activity. This study determined the concentration of stable Cs and Sr, and U in the soil solution, under plant growing conditions. The leafy vegetable komatsuna (Brassica rapa L.) was cultivated for 26 days in pots, where the rhizosphere soil was separated from the non-rhizosphere soil by a nylon net screen. The concentrations of Cs and Sr in the rhizosphere soil solution decreased with time, and were controlled by K+NH(4)(+) and Ca, respectively. On the other hand, the concentration of U in the rhizosphere soil solution increased with time, and was related to the changes of DOC; however, this relationship was different between the rhizosphere and non-rhizosphere soil.

  10. I-129, Kr-85, C-14 and NO/sub x/ removal from spent fuel dissolver off-gas at atmospheric pressure and at reduced off-gas flow

    International Nuclear Information System (INIS)

    Henrich, E.; Huefner, R.

    1981-01-01

    A dissolver off-gas (DOG) system suitable for a LWR, FBR or HTR spent fuel reprocessing plant is described, incorporating the following features: (1) the DOG flow is reduced to a reasonably small volume, using fumeless dissolution conditions, by maintaining high concentrations, the retention procedures are simplified and accompanied by an economic reduction of the equipment size; (2) all process operations are conducted at atmospheric or subatmospheric pressure, including noble gas removal by selective absorption, without using high temperature processes; (3) all processes, except HEPA filtering, are continuous and do not accumulate large amounts of waste nuclides, the DOG process sequence is mutually compatible with itself and with processing in the headend, showing on-line redundancy for the removal of the most radiotoxic nuclides; and (4) the DOG system only deviates slightly from proven technology. The stage of development and relevant results are given both for a lab. scale and a pilot plant scale

  11. Fuel Continuous Mixer ? an Approach Solution to Use Straight Vegetable Oil for Marine Diesel Engines

    OpenAIRE

    Đặng Van Uy; Tran The Nam

    2018-01-01

    The vegetable oil is well known as green fuel for diesel engines due to its low sunphur content and renewable stock. However, there are some problems raising when vegetable oil is used as fuel for diesel engines such as highly effected by cold weather, lower general efficiency, separation in layer if mixed with diesel oil and so on. To overcome that disadvantiges, the authors propose a new idea that to use a continuous fuel mixer to blend vegetable oil with diesel oil to make so called a mixe...

  12. Corrosion of non-irradiated UAl{sub x}-Al fuel in the presence of clay pore solution. A quantitative XRD secondary phase analysis applying the DDM method

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Andreas [Halle-Wittenberg Univ. (Germany). Dept. of Mineralogy and Geochemistry; RWTH Aachen Univ. (Germany). Inst. of Crystallography; Klinkenberg, Martina; Curtius, Hildegard [Forschungszentrum Juelich GmbH (Germany). Inst. of Energy and Climate Research, IEK-6 Nuclear Waste Management

    2017-04-01

    Corrosion experiments with non-irradiated metallic UAl{sub x}-Al research reactor fuel elements were carried out in autoclaves to identify and quantify the corrosion products. Such compounds, considering the long-term safety assessment of final repositories, can interact with the released inventory and this constitutes a sink for radionuclide migration in formation waters. Therefore, the metallic fuel sample was subjected to clay pore solution to investigate its process of disintegration by analyzing the resulting products and the remnants, i.e. the secondary phases. Due to the fast corrosion rate a full sample disintegration was observed within the experimental period of 1 year at 90 C. The obtained solids were subdivided into different grain size fractions and prepared for analysis. The elemental analysis of the suspension showed that, uranium and aluminum are concentrated in the solids, whereas iron was mainly dissolved. Non-ambient X-ray diffraction (XRD) combined with the derivative difference minimization (DDM) method was applied for the qualitative and quantitative phase analysis (QPA) of the secondary phases. Gypsum and hemihydrate (bassanite), residues of non-corroded nuclear fuel, hematite, and goethite were identified. The quantitative phase analysis showed that goethite is the major crystalline phase. The amorphous content exceeded 80 wt% and hosted the uranium. All other compounds were present to a minor content. The obtained results by XRD were well supported by complementary scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) analysis.

  13. Mixture of fuels for solution combustion synthesis of porous Fe{sub 3}O{sub 4} powders

    Energy Technology Data Exchange (ETDEWEB)

    Parnianfar, H.; Masoudpanah, S.M., E-mail: masoodpanah@iust.ac.ir; Alamolhoda, S.; Fathi, H.

    2017-06-15

    Highlights: • Mixture of glycine and urea fuels was applied for solution combustion synthesis of Fe3O4 powders. • The phase and crystallite size of the as-combusted powders depends on the fuel to oxidant ratio (ϕ). • The maximum density (0.033 cm{sup 3}/g) was observed for the as-combusted powders at ϕ = 1. • The highest Ms of 75.5 emu/g and the lowest Hc of 84 Oe were achieved at ϕ = 1. - Abstract: The solution combustion synthesis of porous magnetite (Fe{sub 3}O{sub 4}) powders by a mixture of glycine and urea fuels was investigated concerning the thermodynamic aspects and powder characteristics. The adiabatic combustion temperature and combusted species were thermodynamically calculated as a function of the fuel to oxidant molar ratio (ϕ). The combustion behavior, phase evolution, porous structure and magnetic properties were characterized by thermal analysis, X-ray diffractometry, N{sub 2} adsorption–desorption, electron microscopy and vibrating sample magnetometry techniques. Nearly single phase Fe{sub 3}O{sub 4} powders were synthesized by the mixture of fuels at ϕ values of 0.75 and 1. The as-combusted Fe{sub 3}O{sub 4} powders at ϕ = 1 exhibited porous structure with the specific surface area of 83.4 m{sup 2}/g. The highest saturation magnetization of 75.5 emu/g and the lowest coercivity of 84 Oe were achieved at ϕ = 1, due to the high purity and large crystallite size, inducing from the highest adiabatic combustion temperature.

  14. Innovative automation solutions applied to nuclear fuel production and inspection processes

    International Nuclear Information System (INIS)

    Vas, Ananth

    2012-01-01

    The nuclear industry in India is slated for fast paced growth in the coming years, with a great focus on increasing the capacity for producing, inspecting and finally reprocessing of nuclear fuel. Modern techniques of industrial automation such as robotics, machine vision and laser based systems have been deployed extensively to improve the productivity and output of existing and future installations, particularly for the fuel handling stages mentioned

  15. Simultaneous pollutant removal and electricity generation in denitrifying microbial fuel cell with boric acid-borate buffer solution.

    Science.gov (United States)

    Chen, Gang; Zhang, Shaohui; Li, Meng; Wei, Yan

    2015-01-01

    A double-chamber denitrifying microbial fuel cell (MFC), using boric acid-borate buffer solution as an alternative to phosphate buffer solution, was set up to investigate the influence of buffer solution concentration, temperature and external resistance on electricity generation and pollutant removal efficiency. The result revealed that the denitrifying MFC with boric acid-borate buffer solution was successfully started up in 51 days, with a stable cell voltage of 205.1 ± 1.96 mV at an external resistance of 50 Ω. Higher concentration of buffer solution favored nitrogen removal and electricity generation. The maximum power density of 8.27 W/m(3) net cathodic chamber was obtained at a buffer solution concentration of 100 mmol/L. An increase in temperature benefitted electricity generation and nitrogen removal. A suitable temperature for this denitrifying MFC was suggested to be 25 °C. Decreasing the external resistance favored nitrogen removal and organic matter consumption by exoelectrogens.

  16. A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Berrill, Mark A.; Allu, Srikanth; Hamilton, Steven P.; Sampath, Rahul S.; Clarno, Kevin T. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Dilts, Gary A. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

    2015-04-01

    This paper describes an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors is described. Details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstrating the achieved efficiency of the algorithm are presented. Furthermore, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.

  17. Coal-water fuels - a clean coal solution for Eastern Europe

    International Nuclear Information System (INIS)

    Ljubicic, B.; Willson, W.; Bukurov, Z.; Cvijanovic, P.; Stajner, K.; Popovic, R.

    1993-01-01

    Eastern Europe currently faces great economic and environmental problems. Among these problems is energy provision. Coal reserves are large but cause pollution while oil and gas need to be used for export. Formal 'clean coal technologies' are simply too expensive to be implemented on a large scale in the current economic crisis. The promised western investment and technological help has simply not taken place, western Europe must help eastern Europe with coal technology. The cheapest such technology is coal-water fuel slurry. It can substitute for oil, but research has not been carried out because of low oil prices. Coal-water fuel is one of the best methods of exploiting low rank coal. Many eastern European low rank coals have a low sulfur content, and thus make a good basis for a clean fuel. Italy and Russia are involved in such a venture, the slurry being transported in a pipeline. This technology would enable Russia to exploit Arctic coal reserves, thus freeing oil and gas for export. In Serbia the exploitation of sub-Danube lignite deposits with dredging mining produced a slurry. This led to the use and development of hot water drying, which enabled the removal of many of the salts which cause problems in pulverized fuel combustion. The system is economic, the fuel safer to transport then oil, either by rail or in pipelines. Many eastern European oil facilities could switch. 24 refs

  18. Substitution of chlorinated and fluorinated solvents by biodegradable detergent solution in components cleaning of nuclear fuel elements

    International Nuclear Information System (INIS)

    Vieira, Andre Luiz Pinto da Silva

    2000-01-01

    As the auxiliary oils used in machining evolved from integral into aqueous emulsion, and later on into aqueous-solution synthetic oils, the components cleaning process with organic solvents, originally adopted at the Fuel Element Factory (FEC), Industrias Nucleares do Brasil S.A. (INB) began to present problems in removing oil residues from machined components, due to the incompatibility between aqueous and organic media. In order to eliminate such incompatibility and adapt the process to the environmental laws restricting production and use of chlorinated or fluorinated solvents as a measure for preserving the atmosphere's ozone layer, in 1995 INB initiated the development of a components cleaning process using biodegradable aqueous detergent. The effort was completed in 2000 with the construction of a machine in keeping with the specific geometry of the fuel-assembly components and the operating conditions required for working with the new process. (author)

  19. Geologic disposal as optimal solution of managing the spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Ilie, P.; Didita, L.; Ionescu, A.; Deaconu, V.

    2002-01-01

    To date there exist three alternatives for the concept of geological disposal: 1. storing the high-level waste (HLW) and spent nuclear fuel (SNF) on ground repositories; 2. solutions implying advanced separation processes including partitioning and transmutation (P and T) and eventual disposal in outer space; 3. geological disposal in repositories excavated in rocks. Ground storing seems to be advantageous as it ensures a secure sustainable storing system over many centuries (about 300 years). On the other hand ground storing would be only a postponement in decision making and will be eventually followed by geological disposal. Research in the P and T field is expected to entail a significant reduction of the amount of long-lived radioactive waste although the long term geological disposal will be not eliminated. Having in view the high cost, as well as the diversity of conditions in the countries owning power reactors it appears as a reasonable regional solution of HLW disposal that of sharing a common geological disposal. In Romania legislation concerning of radioactive waste is based on the Law concerning Spent Nuclear Fuel and Radioactive Waste Management in View of Final Disposal. One admits at present that for Romania geological disposal is not yet a stressing issue and hence intermediate ground storing of SNF will allow time for finding a better final solution

  20. Current state of knowledge of water radiolysis effects on spent nuclear fuel corrosion

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    2000-07-01

    Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O 2 or H 2 O 2 in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO 2 in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study , it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed. (author)

  1. Microbial bio-fuels: a solution to carbon emissions and energy crisis.

    Science.gov (United States)

    Kumar, Arun; Kaushal, Sumit; Saraf, Shubhini A; Singh, Jay Shankar

    2018-06-01

    Increasing energy demand, limited fossil fuel resources and climate change have prompted development of alternative sustainable and economical fuel resources such as crop-based bio-ethanol and bio-diesel. However, there is concern over use of arable land that is used for food agriculture for creation of biofuel. Thus, there is a renewed interest in the use of microbes particularly microalgae for bio-fuel production. Microbes such as micro-algae and cyanobacteria that are used for biofuel production also produce other bioactive compounds under stressed conditions. Microbial agents used for biofuel production also produce bioactive compounds with antimicrobial, antiviral, anticoagulant, antioxidant, antifungal, anti-inflammatory and anticancer activity. Because of importance of such high-value compounds in aquaculture and bioremediation, and the potential to reduce carbon emissions and energy security, the biofuels produced by microbial biotechnology might substitute the crop-based bio-ethanol and bio-diesel production.

  2. Fuel reprocessing plant - no solution for the economy of the region

    International Nuclear Information System (INIS)

    Elvers, G.

    1986-01-01

    Both for the construction and operation stage, the direct and indirect impact of the fuel reprocessing plant on employment on the whole will be negative. It is not altogether certain either that there will be no adverse effects for the areas of tourism. The top organization of German trade unions (DGB) holds that a different structure-political concept from the one represented by the large-scale project of the fuel reprocessing plant would be more appropriate for the region. Employment in the steel and construction industries must be safeguarded by corresponding programmes, and new employment must be created in small- and medium-size companies. (DG) [de

  3. Dissolution flowsheet for high flux isotope reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  4. Optimization of fuel cells for BWR using Path Re linking and flexible strategies of solution

    International Nuclear Information System (INIS)

    Castillo M, J. A.; Ortiz S, J. J.; Torres V, M.; Perusquia del Cueto, R.

    2009-10-01

    In this work are presented the obtained preliminary results to design nuclear fuel cells for boiling water reactors (BWR) using new strategies. To carry out the cells design some of the used rules in the fuel administration were discarded and other were implemented. The above-mentioned with the idea of making a comparative analysis between the used rules and those implemented here, under the hypothesis that it can be possible to design nuclear fuel cells without using all the used rules and executing the security restrictions that are imposed in these cases. To evaluate the quality of the obtained cells it was taken into account the power pick factor and the infinite multiplication factor, in the same sense, to evaluate the proposed configurations and to obtain the mentioned parameters was used the CASMO-4 code. To optimize the design it is uses the combinatorial optimization technique named Path Re linking and the Dispersed Search as local search method. The preliminary results show that it is possible to implement new strategies for the cells design of nuclear fuel following new rules. (Author)

  5. Premises and solutions regarding a global approach of gaseous pollutants emissions from the fossil fuel power plants in Romania

    International Nuclear Information System (INIS)

    Tutuianu, O.; Fulger, E.D.; Vieru, A.; Feher, M.

    1996-01-01

    This paper deals with the present state of RENEL (Romanian Electricity Authority) - controlled thermal power plants from the point of view of technical aspects, utilized fuels and pollutant emissions. National and international regulations are also analyzed as well as their implications concerning the management of pollutant atmospheric emissions of the plants of RENEL. Starting from these premises the paper points out the advantage of global approach of pollution problems and offers solutions already implemented by RENEL. This global approach will result in an optimization of costs implied in pollutant emission limitations as the most efficient solution were found and applied. Having in view this treatment of the pollution problems, RENEL has submitted to the Ministry of the Industries and to the Ministry of Waters, Forests and Environmental Protection a 'Convention on the limitation of CO 2 , SO 2 and NO x emissions produced in the thermal power plants of RENEL'. (author)

  6. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Justin H. [BNF Consulting (United States); Kim, Seung Jun, E-mail: skim@lanl.gov [Mechanical and Thermal Engineering Group (AET-1), Los Alamos National Laboratory (United States); Jones, Barclay G. [Department of Nuclear Plasma Radiological Engineering, University of Illinois Urbana-Champaign (United States)

    2016-04-15

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2} can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  7. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    International Nuclear Information System (INIS)

    Joe, Justin H.; Kim, Seung Jun; Jones, Barclay G.

    2016-01-01

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H 2 , O 2 , and H 2 O 2 ) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H 2 , O 2 , and H 2 O 2 ) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H 2 , O 2 , and H 2 O 2 can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  8. Effects of dissolved calcium and magnesium ions on lead-induced stress corrosion cracking susceptibility of nuclear steam generator tubing alloy in high temperature crevice solutions

    International Nuclear Information System (INIS)

    Lu, B.T.; Tian, L.P.; Zhu, R.K.; Luo, J.L.; Lu, Y.C.

    2011-01-01

    The effects of Ca 2+ and Mg 2+ ions on the stress corrosion cracking (SCC) susceptibility of UNS N08800 are investigated using constant extension rate tensile (CERT) tests at 300 o C in simulated crevice chemistries. The presence of lead contamination in the crevice chemistries increases significantly the SCC susceptibility of the alloy. The lead-assisted SCC (PbSCC) susceptibility is reduced markedly by the addition of Ca 2+ and Mg 2+ ions into the solution and this mitigating effect is enhanced by increasing the total concentration of Ca 2+ + Mg 2+ . The CERT test results are consistent with the types of fracture surfaces shown by Scanning Electron Microscopy (SEM). There is a reasonable correlation between the SCC susceptibility and the donor densities in the anodic films in accord with the role of lead-induced passivity degradation in PbSCC.

  9. Coal and wood fuel for electricity production: An environmentally sound solution for waste and demolition wood

    Energy Technology Data Exchange (ETDEWEB)

    Penninks, F.W.M. [EPON, Zwolle (Netherlands)

    1997-12-31

    Waste wood from primary wood processing and demolition presents both a problem and a potential. If disposed in landfills, it consumes large volumes and decays, producing CH{sub 4}, CO{sub 2} and other greenhouse gases. As an energy source used in a coal fired power plant it reduces the consumption of fossil fuels reducing the greenhouse effect significantly. Additional advantages are a reduction of the ash volume and the SO{sub 2} and NO{sub x} emissions. The waste wood requires collection, storage, processing and burning. This paper describes a unique project which is carried out in the Netherlands at EPON`s Gelderland Power Plant (635 MW{sub e}) where 60 000 tonnes of waste and demolition wood will be used annually. Special emphasis is given to the processing of the powdered wood fuel. Therefore, most waste and demolition wood can be converted from an environmental liability to an environmental and economic asset. (author)

  10. Coal and wood fuel for electricity production: An environmentally sound solution for waste and demolition wood

    Energy Technology Data Exchange (ETDEWEB)

    Penninks, F W.M. [EPON, Zwolle (Netherlands)

    1998-12-31

    Waste wood from primary wood processing and demolition presents both a problem and a potential. If disposed in landfills, it consumes large volumes and decays, producing CH{sub 4}, CO{sub 2} and other greenhouse gases. As an energy source used in a coal fired power plant it reduces the consumption of fossil fuels reducing the greenhouse effect significantly. Additional advantages are a reduction of the ash volume and the SO{sub 2} and NO{sub x} emissions. The waste wood requires collection, storage, processing and burning. This paper describes a unique project which is carried out in the Netherlands at EPON`s Gelderland Power Plant (635 MW{sub e}) where 60 000 tonnes of waste and demolition wood will be used annually. Special emphasis is given to the processing of the powdered wood fuel. Therefore, most waste and demolition wood can be converted from an environmental liability to an environmental and economic asset. (author)

  11. Particle Swarm Optimization applied to combinatorial problem aiming the fuel recharge problem solution in a nuclear reactor

    International Nuclear Information System (INIS)

    Meneses, Anderson Alvarenga de Moura; Schirru, Roberto

    2005-01-01

    This work focuses on the usage the Artificial Intelligence technique Particle Swarm Optimization (PSO) to optimize the fuel recharge at a nuclear reactor. This is a combinatorial problem, in which the search of the best feasible solution is done by minimizing a specific objective function. However, in this first moment it is possible to compare the fuel recharge problem with the Traveling Salesman Problem (TSP), since both of them are combinatorial, with one advantage: the evaluation of the TSP objective function is much more simple. Thus, the proposed methods have been applied to two TSPs: Oliver 30 and Rykel 48. In 1995, KENNEDY and EBERHART presented the PSO technique to optimize non-linear continued functions. Recently some PSO models for discrete search spaces have been developed for combinatorial optimization. Although all of them having different formulation from the ones presented here. In this paper, we use the PSO theory associated with to the Random Keys (RK)model, used in some optimizations with Genetic Algorithms. The Particle Swarm Optimization with Random Keys (PSORK) results from this association, which combines PSO and RK. The adaptations and changings in the PSO aim to allow the usage of the PSO at the nuclear fuel recharge. This work shows the PSORK being applied to the proposed combinatorial problem and the obtained results. (author)

  12. Ground-fire effects on the composition of dissolved and total organic matter in forest floor and soil solutions from Scots pine forests in Germany: new insights from solid state 13C NMR analysis

    Science.gov (United States)

    Näthe, Kerstin; Michalzik, Beate; Levia, Delphis; Steffens, Markus

    2016-04-01

    Fires represent an ecosystem disturbance and are recognized to seriously pertubate the nutrient budgets of forested ecosystems. While the effects of fires on chemical, biological, and physical soil properties have been intensively studied, especially in Mediterranean areas and North America, few investigations examined the effects of fire-induced alterations in the water-bound fluxes and the chemical composition of dissolved and particulate organic carbon and nitrogen (DOC, POC, DN, PN). The exclusion of the particulate organic matter fraction (0.45 μm Independent from fire manipulation, the composition of TOM was generally less aromatic (aromaticity index [%] according to Hatcher et al., 1981) with values between 18 (FF) - 25% (B horizon) than the DOM fraction with 23 (FF) - 27% (B horizon). For DOM in FF solution, fire manipulation caused an increase in aromaticity from 23 to 27% compared to the control, due to an increase of the aryl-C and a decrease of the O-alkyl-C and alkyl-C signal. Fire effects were leveled out in the mineral soil. For TOM, fire effects became notable only in the A horizon, exhibiting a decrease in aromaticity from 22 to 18% compared to the control, due to increased O-alkyl-C and diminished aryl-C proportions. Compared to the control, fire only caused minor DOC release rates (events did not significantly enhance the proportion of POC and PN in the total C and N amounts exhibiting values between 10 and 20%. To fully understand the quality and amount of translocated organic C and N compounds within soils under both ambient as well as fire environments, dissolved and particulate size fractions need to be considered.

  13. Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1982-01-01

    The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations

  14. Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

    International Nuclear Information System (INIS)

    Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

    2010-01-01

    Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

  15. Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, K.; Takeuchi, M. [Japan Atomic Energy Agency - JAEA, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

    2016-07-01

    Insoluble sludge is generated in the reprocessing of spent fuel. The sludge obtained from the dissolution of irradiated fuel from the Joyo experimental fast reactor was analyzed to evaluate its chemical form. The sludge was collected by the filtration of the dissolved fuel solution, and then washed in nitric acid. The yields of the sludge weight were less than 1% of the total fuel weight. The chemical composition of the sludge was analyzed after decomposition by alkaline fusion. Molybdenum, technetium, ruthenium, rhodium, and palladium were found to be the main constituent elements of the sludge. X-ray diffraction patterns of the sludge were attributable to Mo{sub 4}Ru{sub 4}RhPd, regardless of the experimental conditions. The concentrations of molybdenum and zirconium in the dissolved fast reactor fuel solutions were low, indicating that zirconium molybdate hydrate (ZMH) is produced in negligible amounts in the process. (authors)

  16. Mechanistic studies of the oxidation of soluble species of ruthenium in nitric acid solutions. Application to the removal of ruthenium from nuclear fuel dissolution solutions

    International Nuclear Information System (INIS)

    Carron, V.

    2001-01-01

    Ruthenium is one of the most troublesome fission products during nuclear fuel reprocessing. His removal from nitric acid fuel dissolution solutions, above the PUREX process, is under consideration. Electro-volatilization could be a possible way to eliminate this element. It consists in the oxidation of soluble ruthenium species coupled with the volatilization of formed RuO 4 . Soluble species are nitrate and nitro complexes of nitrosyl ruthenium RuNO 3+ . The first part of this work deals with the direct oxidation of RuNO 3+ at a golden or a platinum anode. It has been investigated by cyclic voltammetry and infrared and UV-visible reflectance spectroscopy. The oxidation of RuNO 3+ begins with an adsorption step, which precedes the formation of RuO 4 . Then a reaction between RuO 4 and RuNO 3+ occurs to produce a Ru IV compound, which is also electro-oxidized to RuO 4 . The second part concerns potentiostatic electro-volatilization experiences. The rate of electro-volatilization decreases with increasing HNO 3 concentration. At low concentrations, kinetic is controlled by the volatilization of RuO 4 . The rate-determining step is the oxidation of RuNO 3+ at concentrations higher than 1 M. In HNO 3 4 M, the addition of AgNO 3 is required to accelerate the oxidation of RuNO 3+ . The last part is devoted to the study of the indirect oxidation of RuNO 3+ . The electrocatalytic power of electro-generated Ag II is illustrated by voltammetric techniques and potentiostatic electrolysis. The existence of a limit concentration of AgNO 3 is shown (which value depends on experimental conditions) beyond which kinetic is controlled by the RuO 4 volatilization step. These results indicate that the electro-volatilization kinetic could be increased by optimizing the volatilization conditions. (author)

  17. Thermal expansion of TRU nitride solid solutions as fuel materials for transmutation of minor actinides

    International Nuclear Information System (INIS)

    Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo

    2009-01-01

    The lattice thermal expansion of the transuranium nitride solid solutions was measured to investigate the composition dependence. The single-phase solid solution samples of (Np 0.55 Am 0.45 )N, (Pu 0.59 Am 0.41 )N, (Np 0.21 Pu 0.52 Am 0.22 Cm 0.05 )N and (Pu 0.21 Am 0.18 Zr 0.61 )N were prepared by carbothermic nitridation of the respective transuranium dioxides and nitridation of Zr metal through hydride. The lattice parameters were measured by the high temperature X-ray diffraction method from room temperature up to 1478 K. The linear thermal expansion of each sample was determined as a function of temperature. The average thermal expansion coefficients over the temperature range of 293-1273 K for the solid solution samples were 10.1, 11.5, 10.8 and 8.8 x 10 -6 K -1 , respectively. Comparison of these values with those for the constituent nitrides showed that the average thermal expansion coefficients of the solid solution samples could be approximated by the linear mixture rule within the error of 2-3%.

  18. Nuclear fuel technology - Determination of uranium in uranyl nitrate solutions of nuclear grade quality - Gravimetric method

    International Nuclear Information System (INIS)

    2003-01-01

    This International Standard specifies a precise and accurate gravimetric method for determining the mass fraction of uranium in uranyl nitrate solutions of nuclear grade quality containing more than 100 g/kg of uranium. Non-volatile impurities influence the accuracy of the method

  19. Proportioning of 79Se and 126Sn long life radionuclides in the fission products solutions coming from spent fuels processing

    International Nuclear Information System (INIS)

    Comte, J.

    2001-11-01

    The determination of radionuclides present in waste resulting from the nuclear fuel reprocessing is a request from the regulatory authorities to ensure an optimal management of the storage sites. Long-lived radionuclides (T 1/2 > 30 years) are particularly concerned owing to the fact that their impact must be considered for the long term. Safety studies have established a list of long-lived radionuclides (LLRN) whose quantification is essential for the management of the disposal site. Among these, several are pure β emitters, present at low concentration levels in complex matrices. Their determination, by radiochemical method or mass spectrometry, involves selective chemical separations from the others β/γ emitters and from the measurement interfering elements. The work undertaken in this thesis relates to the development of analytical methods for the determination of two long-lived radionuclides: selenium 79 and tin 126, in acid solutions of fission products present in nuclear fuel reprocessing plant. For selenium 79, a β emitter with a half live estimated to be 10 6 years, the bibliography describes different chemical separation methods including precipitation, liquid-liquid extraction and chromatography on ionic resins. After optimisation on a synthetic solution, two of these techniques, precipitation by potassium iodine and separation with ion exchange resins were applied to a genuine solution of fission products at Cogema La Hague. The results showed that only the ion exchange method allows us to obtain a solution sufficiently decontaminated (FDβγ = 250) with a significant selenium recovery yield (85%). This separation allows the measurement of the 79 Se by electrothermal vaporization coupled with inductively coupled plasma mass spectrometry (ETV-ICP/MS), after transfer of the samples to CEA/Cadarache. The concentration of 79 Se measured is 0,42 mg/L in the solution of fission products with an isotopic ratio 79 Se/ 82 Se equal to that recommended by the

  20. Demonstration of an instrumental technique in the measurement of solution weight in the accountability vessels of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nakajima, K.

    1977-04-01

    Load cells were installed on the input accountability vessel of a commercial reactor fuel reprocessing facility to determine if this proven principle of mass measurement is in fact applicable in such a severe radiation environment over a long period of time. Two other locations selected were the plutonium product nitrate solution accountability vessel and the plutonium product nitrate solution storage vessel. The latter two environments, while not severely radio-active, require a high degree of contamination control. All three vessels are of different geometrical configuration and capacity. Each vessel was carefully calibrated for volume measurements by adding controlled pre-measured increments of water. Measurements were made using the conventional dip-tube manometer system and the load cell - digital voltmeter. Standard deviation of the measurements on the input vessel and the plutonium storage vessel were in both cases 0.3%; for the plutonium accountability vessel 1.9%. Measurements taken of the input vessel during the ''cold run'' over a six-month period using solutions of unirradiated uranium showed a standard deviation of 0.4% and a bias of 0.8% in the summer months and 0.7% and 0.6% respectively in the winter months FINAL STOP CODE

  1. A numerical solution model of the rewetting of a nuclear fuel rod

    International Nuclear Information System (INIS)

    Braz Filho, F.A.

    1984-01-01

    The study of thermal behaviour of a nuclear reactor fuel rod during the reflooding phase of the loss-of-coolant accident (LOCA) is presented. A mathematical model and a numerical scheme were proposed in order to solve the bidimensional heat conduction equation in cylindrical coordinates. The phenomenon of reflooding is not completely understood. One of the main difficulties is to estimate the heat transfer coefficient (h). For this reason two different models were elaborated: in the first three regions are considered and in each region h is considered constant; in the second the h profile is adjusted according to the boiling curve. The three region model yields satisfactory results at high and low mass flows while the 'boiling curve' model yields reasonable at low flows. (Author) [pt

  2. Analytical solutions for the temperature field in a 2D incompressible inviscid flow through a channel with walls of solid fuel

    Directory of Open Access Journals (Sweden)

    Sorin BERBENTE

    2011-12-01

    Full Text Available A gas (oxidizer flows between two parallel walls of solid fuel. A combustion is initiated: the solid fuel is vaporized and a diffusive flame occurs. The hot combustion products are submitted both to thermal diffusion and convection. Analytical solutions can be obtained both for the velocity and temperature distributions by considering an equivalent mean temperature where the density and the thermal conductivity are evaluated. The main effects of heat transfer are due to heat convection at the flame. Because the detailed mechanism of the diffusion flame is not introduced the reference chemical reaction is the combustion of premixed fuel with oxidizer in excess. In exchange the analytical solution is used to define an ideal quasi-uniform combustion that could be realized by an n adequate control. The given analytical closed solutions prove themselves flexible enough to adjust the main data of some existing experiments and to suggest new approaches to the problem.

  3. Determination of dissolution rates of spent fuel in carbonate solutions under different redox conditions with a flow-through experiment

    International Nuclear Information System (INIS)

    Roellin, S.; Spahiu, K.; Eklund, U.-B.

    2001-01-01

    Dissolution rates of spent UO 2 fuel have been investigated using flow-through experiments under oxidizing, anoxic and reducing conditions. For oxidizing conditions, approximately congruent dissolution rates were obtained in the pH range 3-9.3 for U, Np, Ba, Tc, Cs, Sr and Rb. For these elements, steady-state conditions were obtained in the flow rate range 0.02-0.3 ml min -1 . The dissolution rates were about 3 mg d -1 m -2 for pH>6. For pH 2 (g) saturated solutions dropped by up to four orders of magnitude as compared to oxidizing conditions. Because of the very low concentrations, only U, Pu, Am, Mo, Tc and Cs could be measured. For anoxic conditions, both the redox potential and dissolution rates increased approaching the same values as under oxidizing conditions

  4. Evaluation of neutronic characteristics of STACY 80-cm-diameter cylindrical core fueled with 6% enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Sono, Hiroki

    2003-06-01

    For the examination of neutronic safety design of forthcoming experimental core configurations in the Static Experiment Critical Facility (STACY), neutronic characteristics of 80-cm-diameter cylindrical cores fueled with 6% enriched uranyl nitrate solution have been evaluated by computational analyses. In the analyses, the latest nuclear data library, JENDL-3.3, was used as neutron cross section data. The neutron diffusion and transport calculations were performed using a diffusion code, CITATION, in the SRAC code system and a continuous-energy Monte Carlo code, MVP. Critical level heights of the cores were obtained using such parameters as uranium concentration (up to 500 gU/l), free nitric acid concentration (up to 8 mol/l), and concentration of soluble neutron poisons, gadolinium and boron. It has been confirmed from the evaluation that all critical cores comply with safety criteria required in the STACY operation concerning excess reactivity, reactivity addition rates and shutdown margins by safety rods. (author)

  5. Solar energised transport solution and customer preferences and opinions about alternative fuel Vehicles – the case of slovenia

    Directory of Open Access Journals (Sweden)

    Matjaž KNEZ

    2015-09-01

    Full Text Available Authorities in Slovenia and other EU member states are confronted with problems of city transportation. Fossil-fuel based transport poses two chief problems – local and global pollution, and dwindling supplies and ever increasing costs. An elegant solution is to gradually replace the present automobile fleet with low emission vehicles. This article first explores the economics and practical viability of the provision of solar electricity for the charging of electric vehicles by installation of economical available PV modules and secondly the customer preferences and opinions about alternative low emission vehicles. Present estimates indicate that for the prevailing solar climate of Celje – a medium-sized Slovenian town – the cost would be only 2.11€ cents/kWh of generated solar electricity. Other results have also revealed that the most relevant factor for purchasing low emission vehicle is total vehicle price.

  6. A fuel-cell reactor for the direct synthesis of hydrogen peroxide alkaline solutions from H(2) and O(2).

    Science.gov (United States)

    Yamanaka, Ichiro; Onisawa, Takeshi; Hashimoto, Toshikazu; Murayama, Toru

    2011-04-18

    The effects of the type of fuel-cell reactors (undivided or divided by cation- and anion-exchange membranes), alkaline electrolytes (LiOH, NaOH, KOH), vapor-grown carbon fiber (VGCF) cathode components (additives: none, activated carbon, Valcan XC72, Black Pearls 2000, Seast-6, and Ketjen Black), and the flow rates of anolyte (0, 1.5, 12 mL h(-1)) and catholyte (0, 12 mL h(-1)) on the formation of hydrogen peroxide were studied. A divided fuel-cell system, O(2) (g)|VGCF-XC72 cathode|2 M NaOH catholyte|cation-exchange membrane (Nafion-117)|Pt/XC72-VGCF anode|2 M NaOH anolyte at 12 mL h(-1) flow|H(2) (g), was effective for the selective formation of hydrogen peroxide, with 130 mA cm(-2) , a 2 M aqueous solution of H(2)O(2)/NaOH, and a current efficiency of 95 % at atmospheric pressure and 298 K. The current and formation rate gradually decreased over a long period of time. The cause of the slow decrease in electrocatalytic performance was revealed and the decrease was stopped by a flow of catholyte. Cyclic voltammetry studies at the VGCF-XC72 electrode indicated that fast diffusion of O(2) from the gas phase to the electrode, and quick desorption of hydrogen peroxide from the electrode to the electrolyte were essential for the efficient formation of solutions of H(2)O(2)/NaOH. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  7. Distribution of 14 elements from two solutions simulating Hanford HLW Tank 102-SY (acid-dissolved sludge and acidified supernate) on four cation exchange resins and five anion exchange resins having different functional groups

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1995-01-01

    As part of the Tank Waste Remediation System program at Los Alamos, we evaluated a series of cation exchange and anion exchange resins for their ability to remove hazardous components from radioactive high-level waste (HLW). The anion exchangers were Reillex TM HPQ, a polyvinyl pyridine resin, and four strong-base polystyrene resins having trimethyl, tri ethyl, tri propyl, and tributyl amine as their respective functional groups. The cation exchange resins included Amberlyst TM 15 and Amberlyst tM XN-1010 with sulfonic acid functionality, Duolite TM C-467 with phosphonic acid functionality, and poly functional Diphonix TM with di phosphonic acid, sulfonic acid, and carboxylic acid functionalities. We measured the distributions of 14 elements on these resins from solutions simulating acid-dissolved sludge (pH 0.6) and acidified supernate (pH 3.5) from underground storage tank 102-SY at the Hanford Reservation near Richland, Washington, USA. To these simulants, we added the appropriate radionuclides and used gamma spectrometry to measure fission products (Ce, Cs, Sr, Tc, and Y), actinides (U, Pu, and Am), and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr). For each of the 252 element/resin/solution combinations, distribution coefficients (Kds) were measured for dynamic contact periods of 30 minutes, 2 hours, and 6 hours to obtain information about sorption kinetics from these complex media. Because we measured the sorption of many different elements, the tabulated results indicate which unwanted elements are most likely to interfere with the sorption of elements of special interest. On the basis of these 756 measured Kd values, we conclude that some of the tested resins appear suitable for partitioning hazardous components from Hanford HLW. (author). 10 refs., 11 tabs

  8. Real time analysis by in line spectrophotometry using optical fibre: application to nuclear fuel reprocessing solutions

    International Nuclear Information System (INIS)

    Pouyat, D.; Couston, L.; Noire, M.H.; Davin, T.; Delage, J.; Bouzon, C.; Goutier, J.; Marty, P.

    1998-01-01

    In nuclear fuel reprocessing factories, an in line determination of actinides and acidity is useful to control the efficiency of the liquid-liquid extraction steps. Although molecular absorptiometric methods are very efficient at the laboratory scale, in-line analysis require to develop passive optical fibre sensor, spectral treatment, and optical fibre active sensors for ions or molecule without optical property such H + . In the first case, a specific optical fibre sensor has been developed to reduce radiological or optical contamination, and to remove the hydraulic perturbations of an intrusive technology. The optical spectrum is directly measured- through a Teflon-PFA tube. Five determination (U IV , U VI , Pu III , Pu IV and HNO 2 ) on eight different process point are achieved every 3 seconds, by using Partial Least Square (PLS) multivariate treatment based on a standards data base. For non linear interference, such as matrix effects on U VI spectrum in nitric acid media, PLS is not very efficient. A physical-chemical model is then required to get a linear relationship. For acidity measurements, an acid-sensitive dye is coated on the core of an optical fiber by the Sol-Gel process. The sensor response, proportional to the indicator protonation, is based on the evanescent wave absorption. This system is free from spectral interference, the response time is fast and measurements are reversible, even with Pu IV at 4 g/l. (author)

  9. Achieving a Green Solution: Limitations and Focus Points for Sustainable Algal Fuels

    Directory of Open Access Journals (Sweden)

    Blanca Antizar-Ladislao

    2012-05-01

    Full Text Available Research investigating the potential of producing biofuels from algae has been enjoying a recent revival due to heightened oil prices, uncertain fossil fuel sources and legislative targets aimed at reducing our contribution to climate change. If the concept is to become a reality however, many obstacles need to be overcome. Recent studies have suggested that open ponds provide the most sustainable means of cultivation infrastructure due to their low energy inputs compared to more energy intensive photobioreactors. Most studies have focused on strains of algae which are capable of yielding high oil concentrations combined with high productivity. Yet it is very difficult to cultivate such strains in open ponds as a result of microbial competition and limited radiation-use efficiency. To improve viability, the use of wastewater has been considered by many researchers as a potential source of nutrients with the added benefit of tertiary water treatment however productivity rates are affected and optimal conditions can be difficult to maintain year round. This paper investigates the process streams which are likely to provide the most viable methods of energy recovery from cultivating and processing algal biomass. The key findings are the importance of a flexible approach which depends upon location of the cultivation ponds and the industry targeted. Additionally this study recommends moving towards technologies producing higher energy recoveries such as pyrolysis or anaerobic digestion as opposed to other studies which focused upon biodiesel production.

  10. Instant release fraction and matrix release of high burn-up UO{sub 2} spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Energy Technology Data Exchange (ETDEWEB)

    Serrano-Purroy, D., E-mail: Daniel.serrano-purroy@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Clarens, F.; Gonzalez-Robles, E. [CTM Centre Tecnologic, Avda. Bases de Manresa 1, 08240 Barcelona (Spain); Glatz, J.P.; Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Pablo, J. de [CTM Centre Tecnologic, Avda. Bases de Manresa 1, 08240 Barcelona (Spain); Department of Chemical Engineering, Universitat Politecnica de Catalunya, Avda. Diagonal 647, 08028 Barcelona (Spain); Casas, I.; Gimenez, J. [Department of Chemical Engineering, Universitat Politecnica de Catalunya, Avda. Diagonal 647, 08028 Barcelona (Spain); Martinez-Esparza, A. [ENRESA, C/Emilio Vargas 7, 28043 Madrid (Spain)

    2012-08-15

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

  11. Liquid scintillation solutions

    International Nuclear Information System (INIS)

    Long, E.C.

    1976-01-01

    The liquid scintillation solution described includes a mixture of: a liquid scintillation solvent, a primary scintillation solute, a secondary scintillation solute, a variety of appreciably different surfactants, and a dissolving and transparency agent. The dissolving and transparency agent is tetrahydrofuran, a cyclic ether. The scintillation solvent is toluene. The primary scintillation solute is PPO, and the secondary scintillation solute is dimethyl POPOP. The variety of appreciably different surfactants is composed of isooctylphenol-polyethoxyethanol and sodium dihexyl sulphosuccinate [fr

  12. Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne

    2006-01-01

    With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge: protection of people and environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, COGEMA LOGISTICS has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear producers through their operational choices. We will focus on the COGEMA LOGISTICS technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfill both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (COGEMA LOGISTICS, TRANSNUCLEAR, MAINCO, and LEMARECHAL CELESTIN) in order to manage transportation of liquid and solid waste towards interim or final storage sites. The paper has the following contents: About radioactive waste; - Radioactive waste classification; - High level activity waste and long-lived intermediate level waste; - Long-lived low level waste; - Short-lived low- and intermediate level waste; - Very low level waste; - The radioactive waste in nuclear fuel cycle; - Packaging design and

  13. Room temperature electrodeposition of actinides from ionic solutions

    Science.gov (United States)

    Hatchett, David W.; Czerwinski, Kenneth R.; Droessler, Janelle; Kinyanjui, John

    2017-04-25

    Uranic and transuranic metals and metal oxides are first dissolved in ozone compositions. The resulting solution in ozone can be further dissolved in ionic liquids to form a second solution. The metals in the second solution are then electrochemically deposited from the second solutions as room temperature ionic liquid (RTIL), tri-methyl-n-butyl ammonium n-bis(trifluoromethansulfonylimide) [Me.sub.3N.sup.nBu][TFSI] providing an alternative non-aqueous system for the extraction and reclamation of actinides from reprocessed fuel materials. Deposition of U metal is achieved using TFSI complexes of U(III) and U(IV) containing the anion common to the RTIL. TFSI complexes of uranium were produced to ensure solubility of the species in the ionic liquid. The methods provide a first measure of the thermodynamic properties of U metal deposition using Uranium complexes with different oxidation states from RTIL solution at room temperature.

  14. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

  15. Removal of actinides from nuclear fuel reprocessing waste solutions with bidentate organophosphorus extractants

    International Nuclear Information System (INIS)

    Schulz, W.W.; McIsaac, L.D.

    1975-08-01

    The neutral bidentate organophosphorus reagents DBDECMP (dibutyl-N,N-diethylcarbamylmethylenephosphonate) and its dihexyl analogue DHDECMP are candidate extractants for removal of actinides from certain acidic waste streams produced at the U. S. ERDA Hanford and Idaho Falls sites. Various chemical and physical properties including availability, cost, purification, alpha radiolysis, and aqueous phase solubility of DBDECMP and DHDECMP are reviewed. A conceptual flowsheet employing a 15 percent DBDECMP (or DHDECMP)--CCl 4 extractant for removal (and recovery) of Am and Pu from Hanford's Plutonium Reclamation Facility acid waste stream (CAW solution) was successfully demonstrated in laboratory-scale mixer-settler tests; this extraction scheme can be used to produce an actinide-free waste. A 30 percent DBDECMP-xylene flowsheet is being tested at the Idaho Falls site for removal of U, Np, Pu, and Am from Idaho Chemical Processing Plant first-cycle high-level raffinate to produce an actinide-free (less than 10 nCi alpha activity/gram) waste. (auth)

  16. Chemical aspects of the precise and accurate determination of uranium and plutonium from nuclear fuel solutions

    International Nuclear Information System (INIS)

    Heinonen, O.J.

    1981-01-01

    A method for the simultaneous or separate determination of uranium and plutonium has been developed. The method is based on the sorption of uranium and plutonium as their chloro complexes on Dowex 1x10 column. When separate uranium and plutonium fractions are desired, plutonium ions are reduced to Pu (III) and eluted, after which the uranium ions are eluted with dilute HCl. Simultaneous stripping of a mass ratio U/Pu approximately 1 fraction for mass spectrometric measurements is achieved by proper choice of eluant HC1 concentration. Special attention was paid to the obtaining of americium free plutonium fractions. The distribution coefficient measurements showed that at 12.5-M HCl at least 30 % of americium ions formed anionic chloro complexes. The chemical aspects of isotopic fractionation in a multiple filament thermal ionization source were also investigated. Samples of uranium were loaded as nitrates, chlorides, and sulphates and the dependence of the measured uranium isotopic ratios on the chemical form of the loading solution as well as on the filament material was studied. Likewise the dependence of the formation of uranium and its oxide ions on various chemical and instrumental conditions was investigated using tungsten and rhenium filaments. Systematic errors arising from the chemical conditions are compared with errors arising from the automatic evaluation of of spectra. (author)

  17. Biamperometric estimation of uranium in input KMP samples of spent fuel reprocessing plant: field experience

    International Nuclear Information System (INIS)

    Gurba, P.B.; Dhakras, S.P.; Chaugule, G.A.; Venugopal, A.K.; Singh, R.K.; Bajpai, D.D.; Nair, P.R.; Xavier, Mary; Aggarwal, S.K.

    2000-01-01

    Feasibility of simple, precise and accurate biamperometric determination of uranium at about 0.1 mg level was earlier established using simulated uranium standards. To evaluate the usefulness of this method for accurate determination of uranium in spent fuel dissolver solution samples, analytical work was carried out

  18. What are the priorities in the fight against fuel poverty? For sustainable solutions, consistent with our commitments. Report no. 18

    International Nuclear Information System (INIS)

    Joly, Guillaume; Guibert, Geraud

    2016-03-01

    In our country, over 5 million households find it hard to pay their energy bills. Heating, lighting and transport are nevertheless essential needs, and one cannot lead a normal life if these needs are not met. Among these households, a certain number find themselves in a genuine situation of fuel poverty. They often rent their homes from the private sector and do not have the means to cover their expenses. These situations present social, but also environmental challenges. In the fight against climate change, the priority must be to eliminate the waste resulting from situations of energy 'leakage'. If sustainable solutions are not found, rising energy prices, which are necessary to reduce greenhouse gas emissions and our carbon footprint, will be unbearably high for a growing number of households and will therefore not be implemented. The fight against fuel poverty has improved in recent years, and new measures formulated by the Energy Transition Law of 2015 are being implemented: energy cheques, energy-saving 'poverty' certificates, etc. But these initiatives are not widely used, they are not well coordinated, and most often they are only calculated on the basis of income. Yet the situation of a household with a low income but living in well-isolated social housing is very different from that of a household living in a home with high energy loss. It is now essential to introduce more clarity and coherence into this system, prioritising long-term solutions which are the only ones capable of reconciling social and environmental priorities. Urgent work needs to be done to tackle energy leakage, particularly in private-sector rental properties, and, as far as possible, we must avoid providing direct aid for fossil fuel use, since this is contrary to the aims of the Paris Climate Change Agreement. Following on from a rigorous and exhaustive assessment, this brief proposes, within the framework of current reforms as well as beyond, to: 1) Fully

  19. Current state of knowledge in radiolysis effects on spent fuel corrosion

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    1998-09-01

    Literature data on the effect of water radiolysis products on spent fuel oxidation and dissolution have been reviewed. Effects of γ-radiolysis, α-radiolysis and dissolved O 2 or H 2 O 2 in unirradiated solutions have been discussed separately. Also the effect of carbonate in γ-irradiated solutions and radiolysis effects on leaching of spent fuels have been reviewed. In addition a radiolysis model for calculation of corrosion rates of UO 2 , presented previously, has been discussed. The model has been shown to give a good agreement between calculated and measured corrosion rates in the case of γ-radiolysis and in unirradiated solutions of dissolved oxygen or hydrogen peroxide. The model has failed to predict the results of α-radiolysis. In a recent study it was shown that the model gave a good agreement with measured corrosion rates of spent fuel exposed in deionized water

  20. Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility

    Energy Technology Data Exchange (ETDEWEB)

    Brueziere, J.; Tribout-Maurizi, A.; Durand, L.; Bertrand, N. [Recycling Business Unit, AREVA, 1 place de la coupole, 92084 Paris La defense Cedex (France)

    2013-07-01

    Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

  1. Reprocessing method of ceramic nuclear fuels in low-melting nitrate molten salts

    International Nuclear Information System (INIS)

    Brambilla, G.; Caporali, G.; Zambianchi, M.

    1976-01-01

    Ceramic nuclear fuel is reprocessed through a method wherein the fuel is dispersed in a molten eutectic mixture of at least two alkali metal nitrates and heated to a temperature in the range between 200 and 300 0 C. That heated mixture is then subjected to the action of a gaseous stream containing nitric acid vapors, preferably in the presence of a catalyst such as sodium fluoride. Dissolved fuel can then be precipitated out of solution in crystalline form by cooling the solution to a temperature only slightly above the melting point of the bath

  2. Effect of Greenhouse Gases Dissolved in Seawater.

    Science.gov (United States)

    Matsunaga, Shigeki

    2015-12-30

    A molecular dynamics simulation has been performed on the greenhouse gases carbon dioxide and methane dissolved in a sodium chloride aqueous solution, as a simple model of seawater. A carbon dioxide molecule is also treated as a hydrogen carbonate ion. The structure, coordination number, diffusion coefficient, shear viscosity, specific heat, and thermal conductivity of the solutions have been discussed. The anomalous behaviors of these properties, especially the negative pressure dependence of thermal conductivity, have been observed in the higher-pressure region.

  3. Theoretical Derivation of Simplified Evaluation Models for the First Peak of a Criticality Accident in Nuclear Fuel Solution

    International Nuclear Information System (INIS)

    Nomura, Yasushi

    2000-01-01

    In a reprocessing facility where nuclear fuel solutions are processed, one could observe a series of power peaks, with the highest peak right after a criticality accident. The criticality alarm system (CAS) is designed to detect the first power peak and warn workers near the reacting material by sounding alarms immediately. Consequently, exposure of the workers would be minimized by an immediate and effective evacuation. Therefore, in the design and installation of a CAS, it is necessary to estimate the magnitude of the first power peak and to set up the threshold point where the CAS initiates the alarm. Furthermore, it is necessary to estimate the level of potential exposure of workers in the case of accidents so as to decide the appropriateness of installing a CAS for a given compartment.A simplified evaluation model to estimate the minimum scale of the first power peak during a criticality accident is derived by theoretical considerations only for use in the design of a CAS to set up the threshold point triggering the alarm signal. Another simplified evaluation model is derived in the same way to estimate the maximum scale of the first power peak for use in judging the appropriateness for installing a CAS. Both models are shown to have adequate margin in predicting the minimum and maximum scale of criticality accidents by comparing their results with French CRiticality occurring ACcidentally (CRAC) experimental data

  4. Antitubercular activity of ZnO nanoparticles prepared by solution combustion synthesis using lemon juice as bio-fuel.

    Science.gov (United States)

    Gopala Krishna, Prashanth; Paduvarahalli Ananthaswamy, Prashanth; Trivedi, Priyanka; Chaturvedi, Vinita; Bhangi Mutta, Nagabhushana; Sannaiah, Ananda; Erra, Amani; Yadavalli, Tejabhiram

    2017-06-01

    In this study, we report the synthesis, structural and morphological characteristics of zinc oxide (ZnO) nanoparticles using solution combustion synthesis method where lemon juice was used as the fuel. In vitro anti-tubercular activity of the synthesized ZnO nanoparticles and their biocompatibility studies, both in vitro and in vivo were carried out. The synthesized nanoparticles showed inhibition of Mycobacterium tuberculosis H37Ra strain at concentrations as low as 12.5μg/mL. In vitro cytotoxicity study performed with normal mammalian cells (L929, 3T3-L1) showed that ZnO nanoparticles are non-toxic with a Selectivity Index (SI) >10. Cytotoxicity performed on two human cancer cell lines DU-145 and Calu-6 indicated the anti-cancer activity of ZnO nanoparticles at varied concentrations. Results of blood hemolysis indicated the biocompatibility of ZnO nanoparticles. Furthermore, in vivo toxicity studies of ZnO nanoparticles conducted on Swiss albino mice (for 14days as per the OECD 423 guidelines) showed no evident toxicity. Copyright © 2017 Elsevier B.V. All rights reserved.

  5. Method to Estimate the Dissolved Air Content in Hydraulic Fluid

    Science.gov (United States)

    Hauser, Daniel M.

    2011-01-01

    In order to verify the air content in hydraulic fluid, an instrument was needed to measure the dissolved air content before the fluid was loaded into the system. The instrument also needed to measure the dissolved air content in situ and in real time during the de-aeration process. The current methods used to measure the dissolved air content require the fluid to be drawn from the hydraulic system, and additional offline laboratory processing time is involved. During laboratory processing, there is a potential for contamination to occur, especially when subsaturated fluid is to be analyzed. A new method measures the amount of dissolved air in hydraulic fluid through the use of a dissolved oxygen meter. The device measures the dissolved air content through an in situ, real-time process that requires no additional offline laboratory processing time. The method utilizes an instrument that measures the partial pressure of oxygen in the hydraulic fluid. By using a standardized calculation procedure that relates the oxygen partial pressure to the volume of dissolved air in solution, the dissolved air content is estimated. The technique employs luminescent quenching technology to determine the partial pressure of oxygen in the hydraulic fluid. An estimated Henry s law coefficient for oxygen and nitrogen in hydraulic fluid is calculated using a standard method to estimate the solubility of gases in lubricants. The amount of dissolved oxygen in the hydraulic fluid is estimated using the Henry s solubility coefficient and the measured partial pressure of oxygen in solution. The amount of dissolved nitrogen that is in solution is estimated by assuming that the ratio of dissolved nitrogen to dissolved oxygen is equal to the ratio of the gas solubility of nitrogen to oxygen at atmospheric pressure and temperature. The technique was performed at atmospheric pressure and room temperature. The technique could be theoretically carried out at higher pressures and elevated

  6. Corrosion studies of thermally sensitised AGR fuel element brace in pH7 and pH9.2 borate solutions

    International Nuclear Information System (INIS)

    Tyfield, S.P.; Smith, C.A.

    1987-04-01

    Brace and cladding of AGR fuel elements sensitised in reactor are susceptible to intergranular and crevice corrosion, which may initiate in the pH7 borate pond storage environment of CEGB/SSEB stations. This report considers the benefit in corrosion control that is provided by raising the pond solution pH to 9.2, whilst maintaining the boron level at 1250 gm -3 . The greater corrosion protection provided by pH9.2 solution compared to the pH7 borate solution is demonstrated by a series of tests with non-active laboratory sensitised brace samples exposed to solutions dosed with chloride or sulphate in order to promote localised corrosion. The corrosion tests undertaken consisted of 5000 hour immersions at 32 0 C and shorter term electrochemically monitored experiments (rest potential, impedance, anodic current) generally conducted at 22 0 C. The pH9.2 solution effectively inhibited the initiation of crevice and intergranular corrosion in the presence of low levels of chloride and sulphate, whereas the pH7 solution did not always do so. However, the pH9.2 solution, dosed with 40 gm -3 chloride, failed to suppress fully crevice corrosion initiated in unborated 40 gm -3 chloride solution at 22 0 C. Fluoride is not deleterious at low levels ∼ 10 gm -3 in the borate solutions. The significant improvement in corrosion control demonstrated for the change from pH7 to pH9.2 borate solution on laboratory sensitised brace samples should ideally be confirmed using complete irradiated AGR fuel elements. (U.K.)

  7. Analyzing Solutions High in Total Dissolved Solids for Rare Earth Elements (REEs) Using Cation Exchange and Online Pre-Concentration with the seaFAST2 Unit; NETL-TRS-7-2017; NETL Technical Report Series; U.S. Department of Energy, National Energy Technology Laboratory: Albany, OR, 2017; p 32

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J. [National Energy Technology Lab. (NETL), Albany, OR (United States); Oregon State Univ., Corvallis, OR (United States). College of Earth, Ocean, and Atmospheric Science; Torres, M. [Oregon State Univ., Corvallis, OR (United States). College of Earth, Ocean, and Atmospheric Science; Verba, C. [National Energy Technology Lab. (NETL), Albany, OR (United States); Oregon State Univ., Corvallis, OR (United States); Hakala, A. [National Energy Technology Lab. (NETL), Pittsburgh, PA, (United States)

    2017-08-01

    The accurate quantification of the rare earth element (REE) dissolved concentrations in natural waters are often inhibited by their low abundances in relation to other dissolved constituents such as alkali, alkaline earth elements, and dissolved solids. The high abundance of these constituents can suppress the overall analytical signal as well as create isobaric interferences on the REEs during analysis. Waters associated with natural gas operations on black shale plays are characterized by high salinities and high total dissolved solids (TDS) contents >150,000 mg/L. Methods used to isolate and quantify dissolved REEs in seawater were adapted in order to develop the capability of analyzing REEs in waters that are high in TDS. First, a synthetic fluid based on geochemical modelling of natural brine formation fluids was created within the Marcellus black shale with a TDS loading of 153,000 mg/L. To this solution, 1,000 ng/mL of REE standards was added based on preliminary analyses of experimental fluids reacted at high pressure and temperature with Marcellus black shale. These synthetic fluids were then run at three different dilution levels of 10, 100, and 1,000–fold dilutions through cation exchange columns using AG50-X8 exchange resin from Eichrom Industries. The eluent from the cation columns were then sent through a seaFAST2 unit directly connected to an inductively coupled plasma mass spectrometer (ICP-MS) to analyze the REEs. Percent recoveries of the REEs ranged from 80–110% and fell within error for the external reference standard used and no signal suppression or isobaric interferences on the REEs were observed. These results demonstrate that a combined use of cation exchange columns and seaFAST2 instrumentation are effective in accurately quantifying the dissolved REEs in fluids that are >150,000 mg/L in TDS and have Ba:Eu ratios in excess of 380,000.

  8. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched 235U fuel pins

    International Nuclear Information System (INIS)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched 235 U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are rather

  9. A study on the alkalimetric titration with gran plot in noncomplexing media for the determination of free acid in spent fuel solutions

    International Nuclear Information System (INIS)

    Suh, Moo Yul; Lee, Chang Heon; Sohn, Se Chul; Kim, Jung Suk; Kim, Won Ho; Eom, Tae Yoon

    1999-01-01

    Based on the study of hydrolysis behaviour of U(VI) ion and major fission product metal ions such as Cs(I), Ce(III), Nd(III), Mo(VI), Ru(II), and Zr(VI) in the titration media, the performance of noncomplexing-alkalimetric titration method for the determination of free acid in the presence of these metal ions was investigated and its results were compared to those from the complexing methods. The free acidities could be determined as low as 0.05 meq in uranium solutions in which the molar ratio of U(VI)/H + was less than 5, when the end-point of titration was estimated by Gran plot. The biases in the determinations were less than ±1% and about +3% respectively for 0.4 meq and 0.05 meq of free acid at the U(VI)/H + molar ratio of up to 5. Applicability of this method to the determination of free acid in spent fuel solutions was confirmed by the analysis of nitric acid content in simulated spent fuel solutions and in a real spent fuel solution

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  11. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  12. Laboratory plant for the separation of cesium from waste solutions of the PUREX process

    International Nuclear Information System (INIS)

    Richter, M.; Eckert, B.; Riemenschneider, J.; Mallon, C.; Mann, D.

    1983-01-01

    A laboratory plant for the separation of cesium from a fission product waste solution of the fuel reprocessing is described. The plant consists of two stages. In the first stage cesium is adsorbed on ammonium molybdatophosphate (AMP). Then the adsorbent is dissolved. From the solution cesium is adsorbed on a cationic ion exchanger in the second stage. Then AMP can be reproduced from this solution. For the elution of cesium in the second stage a NH 4 NO 3 solution (3 m) is used. Flow sheet, construction and the control device of the plant are described and the results of tests with a model solution are given. (author)

  13. Monte Carlo criticality analysis for dissolvers with neutron poison

    International Nuclear Information System (INIS)

    Yu, Deshun; Dong, Xiufang; Pu, Fuxiang.

    1987-01-01

    Criticality analysis for dissolvers with neutron poison is given on the basis of Monte Carlo method. In Monte Carlo calculations of thermal neutron group parameters for fuel pieces, neutron transport length is determined in terms of maximum cross section approach. A set of related effective multiplication factors (K eff ) are calculated by Monte Carlo method for the three cases. Related numerical results are quite useful for the design and operation of this kind of dissolver in the criticality safety analysis. (author)

  14. Sogin enriched uranium extraction (EUREX) plant spent fuel pool cleaning and decontamination utilizing the Smart Safe solution

    International Nuclear Information System (INIS)

    Denton, M.S.; Gili, M.; Nasta, M.; Quintiliani, R.; Caccia, G.; Botzen, W.; Forrester, K.

    2009-01-01

    SOGIN's EUREX facility in Italy was developed as a pilot plant functional testing laboratory for spent fuel reprocessing. This facility was operated successfully for many years since 1970 and was eventually shutdown consistent with Italy's suspension of all nuclear operations. At the time of suspension, the EUREX facility still had spent nuclear fuel assemblies in storage from a nearby PWR. Other fuel assemblies from an Italian AGR had remained stored in the spent fuel pool for the 20 years or so waiting for removal and reprocessing abroad. Being Magnox fuel elements, their recovery for the transport produced a huge amount of sludge in the pool. During this time, sediment, dirt, corrosion products, fuel cladding, etc. has collected within the fuel pool as a crud layer dispersed throughout. Most of this crud has accumulated on the horizontal surfaces of the pool and fuel element assemblies, while some remains as a suspended colloidal material. Furthermore many other contaminated metal components, used during the operation years, where still inside the pool. Due to a pool leak discovered in 2006, SOGIN speeded up its pool decommissioning program, making also available the transfer of the spent fuel to a nearby interim repository, with the goal to completely drain the pool in the shortest period of time. In order for SOGIN to successfully transfer the fuel assemblies from their current storage basket locations to the spent fuel transfer cask, the bulk of the crud needed to be removed. This cleanup operation was deemed necessary to minimize the suspension of contamination in the water during underwater handling operations. This would reduce the decontamination efforts on the transfer cask upon removal, once loaded with the spent fuel, and enhance safety by reducing potential underwater visibility issues. The operations were completed in July 2008 with the release to the environment of the pool water, thoroughly purified and without any relevant radiological impact. The

  15. Solution of the two dimensional diffusion and transport equations in a rectangular lattice with an elliptical fuel element using Fourier transform methods: One and two group cases

    International Nuclear Information System (INIS)

    Williams, M.M.R.; Hall, S.K.; Eaton, M.D.

    2014-01-01

    Highlights: • A rectangular reactor cell with an elliptical fuel element. • Solution of transport and diffusion equations by Fourier expansion. • Numerical examples showing convergence. • Two group cell problems. - Abstract: A method for solving the diffusion and transport equations in a rectangular lattice cell with an elliptical fuel element has been developed using a Fourier expansion of the neutron flux. The method is applied to a one group model with a source in the moderator. The cell flux is obtained and also the associated disadvantage factor. In addition to the one speed case, we also consider the two group equations in the cell which now become an eigenvalue problem for the lattice multiplication factor. The method of solution relies upon an efficient procedure to solve a large set of simultaneous linear equations and for this we use the IMSL library routines. Our method is compared with the results from a finite element code. The main drawback of the problem arises from the very large number of terms required in the Fourier series which taxes the storage and speed of the computer. Nevertheless, useful solutions are obtained in geometries that would normally require the use of finite element or analogous methods, for this reason the Fourier method is useful for comparison with that type of numerical approach. Extension of the method to more intricate fuel shapes, such as stars and cruciforms as well as superpositions of these, is possible

  16. Method and Result of Experiment for Support of Technical Solutions in the Field of Perfection of a Nuclear Fuel Cycle for Future PWR Reactors

    International Nuclear Information System (INIS)

    Ostrovskiy, V.; Kudryavtsev, E.; Tutnov, I.

    2011-01-01

    The paper presents the basics of approach of planning and carrying out of experiments to validate safety PWR reactors of the future when accepting technical solutions concerning using of improved fuel rods in fuel assembly. Basic principles and criteria used for the validation of technical solutions and developments in improving of nuclear fuel cycle of PWR reactors of the future are presented from the point of safety of future operation of modified fuel rods. We explore the questions of safety operation of PWR reactors with fuel assemblies, containing fuel rods with different length of fuel. The paper discusses the ways of solving of important tasks of critical facility experiments conducting for verification of new technical solutions in the sphere of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations of IEC (International Electromechanical Commission). New Federal laws of Russian Federation define the main principle for demands to NPP and any supplier of nuclear techniques. The principle is 'quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety for any moment of operation of NPP'. On the other hand the second principle should be applied to extraction of the greatest benefit from operation of the equipment, systems or the NPP as whole: 'The long operation and full commercial use of resource and service properties of the equipment, systems and the NPP as a whole'. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components. Solving the practical problems of a validation of safety use of fuel rods with the increased length of a fuel column in fuel assembly in nuclear reactors of the future, we should choose new strategies and programs of verification experiments on the base of the analysis of guarantees

  17. Criticality safety of the ten-well insert for the pot dissolver

    International Nuclear Information System (INIS)

    Forstner, J.L.

    1982-05-01

    Nuclear safety for most fuels dissolved at SRP is ensured by some form of insert with a favorable geometry in a pot dissolver. A ten-well insert was designed which would permit an adequate charge of highly enriched U-Al alloy fuels of the MTR type. It can handle cylindrical fuel bundles up to 5 in. dia. Dependence on administrative control is reduced. 10 figures

  18. Biogasoline: An out-of-the-box solution to the food-for-fuel and land-use competitions

    International Nuclear Information System (INIS)

    Hassan, S.N.; Sani, Y.M.; Abdul Aziz, A.R.; Sulaiman, N.M.N.; Daud, W.M.A.W.

    2015-01-01

    Highlights: • Reviewed prospects of biogasoline production as alternative to ethanol and biodiesel. • Biogasoline promises to be cheaper and more environmental friendly. • Inedible feedstocks would ensure higher net energy gain. • Inedible feedstocks will resolve food-for-fuel conflicts and land-use competitions. • Advances required for producing bioenergy crop and renewable energy sources. - Abstract: Societal developments are hinged on the energy supplied by fossil fuels. However, the supply of these fuels is finite in the foreseeable future. This is aside the associated environmental degradation and economic sustainability of these fuels. These negative consequences and challenges spurred the search for sustainable energy sources such as biofuels. However, affordable feedstocks and efficient synthesis for renewable fuels remain indispensable and yet challenging line of research. Therefore, breakthroughs in plant biotechnology and mass production are essential prerequisites for ensuring the sustainability of biofuels as alternatives to petroleum-based energy. Conversely, public outcry concerning the food-for-fuel conflicts and land-use change hinder the popularity of such biofuel energy sources. Therefore, this paper reviewed the prospects of biogasoline production as sustainable alternative to ethanol and a compliment to biodiesel. Apart from reduction in greenhouse gas emissions, biogasoline promises to be cheaper and more environmental friendly. Further, inedible feedstocks such as microalgae and rubber seed oil would ensure higher net energy gain. Consequently, these will help resolve the food-for-fuel conflicts and land-use competitions. However, achieving the biofuel central policy depends on advances in processing the renewable energy sources

  19. Effect of diluent wash over the removal of aqueous dissolved TBP and DBP in reprocessing

    International Nuclear Information System (INIS)

    Manjula, R.; Dasi, Mahesh; Mohandas, Jaya; Vijaya Kumar, N.; Kumar, T.

    2015-01-01

    In reprocessing of nuclear spent fuels by PUREX process Tri-n-Butyl phosphate diluted with n-Dodecane (nDD) is used as solvent. This solvent undergoes degradation due to radiation yielding degradation products, mainly Di-n-butyl phosphate (HDBP). During extraction steps some amount of these organic gets dissolved in aqueous phase owing to its mutual solubility. Removal of dissolved organic from aqueous streams before evaporation is essential to prevent red oil related disasters. Diluent wash technique employing nDD as diluent is one of the commonly used method for the same. During the continuous operation of this process, the diluent will get loaded with dissolved organic and subsequently the performance of diluent will not remain same as pure diluent. While some reports are available in literature for the efficiency of removal of TBP by nDD, so far no work has been reported for the removal of DBP. The scope of the present work is to ascertain the efficiency of diluent wash technique on the removal of dissolved TBP as well as DBP. The results obtained indicate that the removal of dissolved TBP by nDD decreases with increase in percentage of TBP in nDD. In the case of DBP it is just reverse and the removal becomes more effective when the TBP percentage in the diluent increases. A/O ratio of 6:1 is found to be more suitable. As the DBP is getting extracted very effectively into nDD containing TBP, diluent wash solution should be treated as spent organic and managed accordingly for further utilization

  20. LARGE-scale forest fuel supply solution trough a regional terminal network; Terminaalitoimintoihin perustuvan metsaepolttoaineen hankintalogistiikkajaerjestelmaen kehittaeminen

    Energy Technology Data Exchange (ETDEWEB)

    Leppaenen, T. [Etelae-Savon Energia Oy, Mikkeli (Finland)

    2006-12-19

    The aim of the study is to develop logistic systems for supply of forest fuel where a terminal is part of the supply chain. Operations in the terminal, supply chains of the forest fuel and joining them to the terminal network are testing and following p. Also operation and business models are under analyzing. Costs, cost factors, benefits and space requirement of the terminal and cost-effectiveness of the entrepreneurship of the terminal are carried out. (orig.)

  1. Experimental determination and chemical modelling of radiolytic processes at the spent fuel/water interface. Experiments carried out in carbonate solutions in absence and presence of chloride

    Energy Technology Data Exchange (ETDEWEB)

    Bruno, Jordi; Cera, Esther; Grive, Mireia; Duro, Lara [Enviros Spain SL (Spain); Eriksen, Trygve [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Nuclear Chemistry

    2003-01-01

    We report on the recent experimental and modelling results of a research programme that started in 1995. The aim has been to understand the kinetic and thermodynamic processes that control the radiolytic generation of oxidants and reductants at the spent fuel water interface and their consequences for spent fuel matrix stability and radionuclide release. This has been done by carrying out well-controlled dissolution experiments of PWR Ringhals spent fuel fragments in an initially anoxic closed system and by using different solution compositions. Experimental series started with several tests carried out with deionised water as solvent, in a second phase experiments were conducted with 10 mM bicarbonate solutions. New experimental series were set up during the last two years by using the same bicarbonate content in solutions with varying NaCl concentrations in order to ascertain the role of this ligand on the radiolytic products and its consequence for radionuclide release. The selected NaCl concentrations are in the range of 0.1 to 10 mM. Experimental data shows that uranium dissolution at early contact times is controlled by the oxidation of the UO{sub 2} matrix. This process controls the co-dissolution of most of the analysed radionuclides, including Sr, Mo, Tc, Np and surprisingly enough, Cs. In the overall the release rates for U and the matrix associated radionuclides are in the range of 10{sup -6} moles/day with a clear decreasing trend with exposure time and after 2 years the initial release rates have decreased down to 3x10{sup -8} moles/day. The solubility of the released actinides appears to be limited by the formation of An(IV) hydroxide phases, although Np concentrations in solution did not reach solubility levels during the time intervals of the present tests. No secondary solid phase appears to control the solubility of the rest of the elements.

  2. Fuel Chemistry Division: progress report for 1985

    International Nuclear Information System (INIS)

    1988-01-01

    Fuel Chemistry Division was formed in May 1985 to give a larger emphasis on the research and development in chemistry of the nuclear fuel cycle. The areas of research in Fuel Chemistry Division are fuel development and its chemical quality control, understanding of the fuel behaviour and post irradiation examinations, chemistry of reprocessing and waste management processes as also the basic aspects of actinide and relevant fission product elements. This report summarises the work by the staff of the Division during 1985 and also some work from the previous periods which was not reported in the progress reports of the Radiochemistry Division. The work related to the FBTR fuel was one of the highlights during this period. In the area of process chemistry useful work has been carried out for processing of plutonium bearing solutions. In the area of mass spectrometry, the determination of trace constituents by spark source mass spectrometry has been a major area of research. Significant progress has also been made in the use of alpha spectromet ry techniques for the determination of plutonium in dissolver solution and other samples. The technology of plutonium utilisation is quite complex and the Division would continue to look into the chemical aspects of this technology and provide the necessary base for future developments in this area. (author)

  3. Selective extraction of metals from acidic uranium(VI) solutions using neo-tridecano-hydroxamic acid

    International Nuclear Information System (INIS)

    Bardoncelli, F.; Grossi, G.

    1975-01-01

    According to this invention neo-alkyl-hydroxamic acids are employed as ion-exchanging agents in processes for liquid-liquid extraction with the aim of separating, purifying dissolved metals and of converting a metal salt solution into a solution of a salt of the same metal but with different anion. In particular it is an objective of this invention to provide a method whereby a molecular pure uranium solution is obtained by selective extraction from a uranium solution delivered by irradiated fuel reprocessing plants and containing plutonium, fission products and other unwanted metals, in which method neo-tridecane-hydroxamic acid is employed as ion exchanger. (Official Gazette)

  4. Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions

    Energy Technology Data Exchange (ETDEWEB)

    Mousavi Shirazi, Seyed Alireza [Islamic Azad Univ. (I.A.U.), Dept. of Physics, Tehran (Iran, Islamic Republic of)

    2015-07-15

    In this symbolic investigation, a cylindrical cell in a LWR, which consists of one fuel pin and moderator (water), is considered. The width of this cylindrical cell is divided into 100 equal units. Since the neutron flux in a cylindrical fuel pin is resulting from the diffusion equation: -(1)/(r)(d)/(dr)Dr(d)/(dr)φ(r) + Σ{sub a}φ(r) = S(r), the amount of fast neutron fluxes are obtained on the basis of the numeric solution of this equation, and the applied boundary conditions are considered: φ'(0) = φ'(1) = 0. This differential equation is solved by the tridiagonal method for variant enrichments of uranium. Neutron fluxes are obtained in variant radii of fuel pin and moderator and are finally compared with each other. There are some interesting outcomes resulting from this investigation. It can be inferred that because of the fuel enrichment increment, the fast neutron flux increases significantly at the centre of core, while many of the fast neutrons produced are absorbed after entering the water region, moderation of lots of them causes the reduced neutron flux to get improved in this region.

  5. Studies on PEM fuel cell noble metal catalyst dissolution

    DEFF Research Database (Denmark)

    Andersen, S. M.; Grahl-Madsen, L.; Skou, E. M.

    2011-01-01

    A combination of electrochemical, spectroscopic and gravimetric methods was carried out on Proton Exchange Membrane (PEM) fuel cell electrodes with the focus on platinum and ruthenium catalysts dissolution, and the membrane degradation. In cyclic voltammetry (CV) experiments, the noble metals were...... found to dissolve in 1 M sulfuric acid solution and the dissolution increased exponentially with the upper potential limit (UPL) between 0.6 and 1.6 vs. RHE. 2-20% of the Pt (depending on the catalyst type) was found to be dissolved during the experiments. Under the same conditions, 30-100% of the Ru...... (depending on the catalyst type) was found to be dissolved. The faster dissolution of ruthenium compared to platinum in the alloy type catalysts was also confirmed by X-ray diffraction measurements. The dissolution of the carbon supported catalyst was found one order of magnitude higher than the unsupported...

  6. Investigation of the gas formation in dissolution process of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Zhang Qinfen; Liao Yuanzhong; Chen Yongqing; Sun Shuyun; Fan Yincheng

    1987-12-01

    The gas formation in dissolution process of two kinds of nuclear fuels was studied. The results shows that the maximum volume flow released from dissolution system is composed of two parts. One of them is air remained in dissolver and pushed out by acid vapor. The other is produced in dissolution reaction. The procedure of calculating the gas amount produced in dissolution process has been given. It is based on variation of components of dissolution solution. The gas amount produced in dissolution process of spent UO 2 fuel elements was calculated. The condenser system and loading volume of disposal system of tail gas of dissolution of spent fuel were discussed

  7. Development of a recovery process of scraps resulting from the manufacture of metallic uranium fuels

    International Nuclear Information System (INIS)

    Camilo, Ruth L.; Kuada, Terezinha A.; Forbicini, Christina A.L.G.O.; Cohen, Victor H.; Araujo, Bertha F.; Lobao, Afonso S.T.

    1996-01-01

    The study of the dissolution of natural metallic uranium fuel samples with aluminium cladding is presented, in order to obtain optimized conditions for the system. The aluminium cladding was dissolved in an alkaline solution of Na OH/Na NO 3 and the metallic uranium with HNO 3 . A fumeless dissolution with total recovery of nitrous gases was achieved. The main purpose of this project was the recovery of uranium from scraps resulting from the manufacture of the metallic uranium fuel or other non specified fuels. (author)

  8. New process of the preparation of catalyzed gas diffusion electrode for PEM fuel cells based on ultrasonic direct solution spray reaction method

    Energy Technology Data Exchange (ETDEWEB)

    Oishi, K.; Savadogo, O. [Ecole Polytechnique de Montreal, Montreal, PQ (Canada). Laboratoire de nouveaux materiaux pour l' energie et l' electrochimie

    2008-07-01

    This paper reported on a newly developed process for in-situ catalyst deposition on gas diffusion electrodes (GDE) for polymer electrolyte fuel cells. This process has the potential to reduce the number of steps for catalyzed GDE fabrication. In addition, the process offers economic advantages for the fuel cell commercialization. In this study, a home-made catalyst maker with ultrasonic spray method was used to prepare a solution of the carbon supported platinum catalyst on the GDL. The sprayed catalyst powder consisted of carbon support. The catalyst particles did not prevent gas flow channels on the GDL. The catalyst layer was shown to be located only on the top surface of the GDL and was not packed into its flow channel. Results of Cross-section SEM image, crystallization, micro structure and electro-catalytic activity for the oxygen reduction reaction were also discussed. 1 ref., 1 fig.

  9. Effect of Greenhouse Gases Dissolved in Seawater

    Directory of Open Access Journals (Sweden)

    Shigeki Matsunaga

    2015-12-01

    Full Text Available A molecular dynamics simulation has been performed on the greenhouse gases carbon dioxide and methane dissolved in a sodium chloride aqueous solution, as a simple model of seawater. A carbon dioxide molecule is also treated as a hydrogen carbonate ion. The structure, coordination number, diffusion coefficient, shear viscosity, specific heat, and thermal conductivity of the solutions have been discussed. The anomalous behaviors of these properties, especially the negative pressure dependence of thermal conductivity, have been observed in the higher-pressure region.

  10. The fission products palladium and rhodium: Their state in solutions, their behavior in the regeneration of fuel of atomic power stations, and the search for selective extraction techniques

    International Nuclear Information System (INIS)

    Arseenkov, L.V.; Zakharkin, B.S.; Lunichkina, K.P.; Renard, E.V.; Rogozhkin, V.Yu.; Shorokhov, N.A.

    1992-01-01

    At the present time many research centers are working on the extraction of noble metals in the form of fission fragments. Consistent data has been obtained on the mass accumulation of noble metals in various forms of processed nuclear fuel. Requirements are noted that must be met for obtaining industrial and economic efficiency in the extraction of noble metals by the Purex process. Presently there is a lack of information on the extraction of noble metals from spent fuel, particularly as far as the nitric acid media of the Purex process are concerned. The authors will discuss individual test observations on simulating systems and real systems with noble metals. The investigations focused on the noble metals of lowest radioactivity, namely palladium and rhodium. The complexity of the chemistry of ruthenium, on the one hand, and the possible selective, clearing distillation of ruthenium tetroxide from nitric acid solutions, on the other hand, make it necessary to focus the attention on the unresolved problems of the extraction of palladium and rhodium. The article further includes discussion on the following topics: noble metals in solutions of purex process, electrochemical operations involving noble metals, extraction systems for rhodium and palladium, separation of palladium from real solutions

  11. Effects of Fuel to Synthesis of CaTiO3 by Solution Combustion Synthesis for High-Level Nuclear Waste Ceramics.

    Science.gov (United States)

    Jung, Choong-Hwan; Kim, Yeon-Ku; Han, Young-Min; Lee, Sang-Jin

    2016-02-01

    A solution combustion process for the synthesis of perovskite (CaTiO3) powders is described. Perovskite is one of the crystalline host matrics for the disposal of high-level radioactive wastes (HLW) because it immobilizes Sr and Lns elements by forming solid solutions. Solution combustion synthesis, which is a self-sustaining oxi-reduction reaction between nitrate and organic fuel, the exothermic reaction, and the heat evolved convert the precursors into their corresponding oxide products above 1100 degrees C in air. To investigate the effects of amino acid on the combustion reaction, various types of fuels were used; a glycine, amine and carboxylic ligand mixture. Sr, La and Gd-nitrate with equivalent amounts of up to 20% of CaTiO3 were mixed with Ca and Ti nitrate and amino acid. X-ray diffraction analysis, SEM and TEM were conducted to confirm the formed phases and morphologies. While powders with an uncontrolled shape are obtained through a general oxide-route process, Ca(Sr, Lns)TiO3 powders with micro-sized soft agglomerates consisting of nano-sized primary particles can be prepared using this method.

  12. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

  13. Process automation using combinations of process and machine control technologies with application to a continuous dissolver

    International Nuclear Information System (INIS)

    Spencer, B.B.; Yarbro, O.O.

    1991-01-01

    Operation of a continuous rotary dissolver, designed to leach uranium-plutonium fuel from chopped sections of reactor fuel cladding using nitric acid, has been automated. The dissolver is a partly continuous, partly batch process that interfaces at both ends with batchwise processes, thereby requiring synchronization of certain operations. Liquid acid is fed and flows through the dissolver continuously, whereas chopped fuel elements are fed to the dissolver in small batches and move through the compartments of the dissolver stagewise. Sequential logic (or machine control) techniques are used to control discrete activities such as the sequencing of isolation valves. Feedback control is used to control acid flowrates and temperatures. Expert systems technology is used for on-line material balances and diagnostics of process operation. 1 ref., 3 figs

  14. Remote repair of the dissolvers in Tokai reprocessing plant

    International Nuclear Information System (INIS)

    Otani, Yosikuni

    1985-01-01

    In the Tokai fuel reprocessing plant, there occurred failures (pinholes) in two dissolver tanks successively in 1982 and 1983. These dissolvers are set under high radiation field, not permitting access of the personnel. So, repair works were carried out after development of the remotely operated repair system. For repair of the failed dissolver tanks, after tests and studies, the means was employed of grinding off the wall surface to small depth and then forming over it a corrosion resistant sealing layer by padding welding. The repair system which enabled the repair and the inspection in the cell by remote operation consisted of six devices including polishing, welding, dye penetration test, etc. Repair works on the dissolvers took two months and a half from September 1983. (Mori, K.)

  15. Research problems of fission product behaviour in fuels of nuclear power plants and ways of their solution

    International Nuclear Information System (INIS)

    Sulaberidze, V.Sh.

    1988-01-01

    The most important problems of studying behaviour of fission products in fuel elements of maneouvrable nuclear power plants units are formulated. In-pile and out-of-pile investigation methods solving these problems are characterized in brief. 12 refs.; 2 figs

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO 2 fuel dissolved in nitric acid and UO 2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. - Highlights: • The noble metal phase was chemically extracted from spent nuclear fuel and analyzed non-destructively. • Noble metal phase nuclides and long-lived iodine were identified and quantified using neutron activation analysis. • Activation to shorter-lived radionuclides allowed rapid analysis of long-lived fission products in spent fuel using gamma spectrometry

  17. 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)

  18. Release of Dissolved CO2 from Water in Laboratory Porous Media Following Rapid Depressurization

    Science.gov (United States)

    Crews, J. B.; Cooper, C. A.

    2011-12-01

    A bench-top laboratory study is undertaken to investigate the effects of seismic shocks on brine aquifers into which carbon dioxide has been injected for permanent storage. Long-term storage in deep saline aquifers has been proposed and studied as one of the most viable near-term options for sequestering fossil fuel-derived carbon dioxide from the atmosphere to curb anthropogenic climate change. Upon injection into the subsurface, it is expected that CO2, as either a gas or supercritical fluid, will mix convectively with the formation water. The possibility exists, however, that dissolved CO2 will come out of solution as a result of an earthquake. The effect is similar to that of slamming an unsealed container of carbonated beverage on a table; previously dissolved CO2 precipitates, forms bubbles, and rises due to buoyancy. In this study, we measure the change in gas-phase CO2 concentration as a function of the magnitude of the shock and the initial concentration of CO2. In addition, we investigate and seek to characterize the nucleation and transport of CO2 bubbles in a porous medium after a seismic shock. Experiments are conducted using a Hele-Shaw cell and a CCD camera to quantify the fraction of dissolved CO2 that comes out of solution as a result of a sharp mechanical impulse. The data are used to identify and constrain the conditions under which CO2 comes out of solution and, further, to understand the end-behavior of the precipitated gas-phase CO2 as it moves through or is immobilized in a porous medium.

  19. Liquid scintillation solution

    International Nuclear Information System (INIS)

    Long, E.C.

    1977-01-01

    A liquid scintillation solution is described which includes (1) a scintillation solvent (toluene and xylene), (2) a primary scintillation solute (PPO and Butyl PBD), (3) a secondary scintillation solute (POPOP and Dimethyl POPOP), (4) a plurality of substantially different surfactants and (5) a filter dissolving and/or transparentizing agent. 8 claims

  20. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 2. of ISO 7097 describes procedures for determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with cerium(IV) and ISO 7097-1 uses a titration with potassium dichromate

  1. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 1. of ISO 7097 describes procedures for the determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with potassium dichromate and ISO 7097-2 uses a titration with cerium(IV)

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

    Science.gov (United States)

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

    2015-07-01

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

  3. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  4. An instrument for the examination of nucleation from solution and its applications to the study of precipitation from diesel fuels and solutions of n-alkanes

    Energy Technology Data Exchange (ETDEWEB)

    Gerson, A R; Roberts, K J; Sherwood, J N [Strathclyde Univ., Glasgow (UK). Dept. of Pure and Applied Chemistry

    1991-03-01

    An automated apparatus has been designed and constructed to measure precipitation and dissolution of solids from solution at varying automatically predetermined rates of heating and cooling. The appearance and disappearance of crystals are detected by means of a fibre optic turbidity sensor attached to a Sybron/Brinkman colorimeter. Temperature is measured by a Pt resistance thermometer attached to a constant current source. Both of these measurements are recorded by a personal computer via an analog to digital converter. The temperature of the system and its variation is controlled from the personal computer via a digital to analog interface attached to the control head of a Haake F3Q cryostat. The system has been used for measurements of precipitation and dissolution temperatures for diesel waxes with and without nucleation additives. Studies have been made of the nucleation of a single n-alkane from solution. From these measurements, saturation curves, orders of reactions, interfacial tensions and the critical radii of nuclei can be assessed. (orig.).

  5. The radiolysis of solutions containing Pu(6)

    International Nuclear Information System (INIS)

    Rance, P.J.W.; Zilberman, B.Y.

    2000-01-01

    The reduction of Pu(VI) in nitric acid solutions containing uranium and various fission product elements as a result of both its inherent alpha radiation and also external gamma irradiation at dose rates similar to those experienced by dissolved fuel solutions has been investigated. The presence of the additional metals has been shown to eliminate the induction periods required prior to the reduction of Pu(VI) in nitric acid. G values for the auto-radiolytic reduction of Pu(VI) have been found to be between 0.6 and 1.1 for 3 g/1 Pu solutions containing between 0.12 and 9.2 % 238 Pu (balance 239 Pu). Uranium and palladium have been found to accelerate the reduction of Pu(VI) during gamma irradiation at dose rates of between 0.41 and 1.64 kGy/hour. (authors)

  6. The radiolysis of solutions containing Pu(6)

    Energy Technology Data Exchange (ETDEWEB)

    Rance, P.J.W. [BNFL British Nuclear Fuels, Sellafield, Seascale, Cumbria, Research and Technology (United Kingdom); Zilberman, B.Y. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

    2000-07-01

    The reduction of Pu(VI) in nitric acid solutions containing uranium and various fission product elements as a result of both its inherent alpha radiation and also external gamma irradiation at dose rates similar to those experienced by dissolved fuel solutions has been investigated. The presence of the additional metals has been shown to eliminate the induction periods required prior to the reduction of Pu(VI) in nitric acid. G values for the auto-radiolytic reduction of Pu(VI) have been found to be between 0.6 and 1.1 for 3 g/1 Pu solutions containing between 0.12 and 9.2 % {sup 238}Pu (balance {sup 239}Pu). Uranium and palladium have been found to accelerate the reduction of Pu(VI) during gamma irradiation at dose rates of between 0.41 and 1.64 kGy/hour. (authors)

  7. On the observation of a huge lattice contraction and crystal habit modifications in LiMn2O4 prepared by a fuel assisted solution combustion

    International Nuclear Information System (INIS)

    Ragavendran, K.; Sherwood, D.; Vasudevan, D.; Emmanuel, Bosco

    2009-01-01

    Two batches of poly-crystalline lithium manganate were prepared by a fuel assisted solution combustion method. LiMn 2 O 4 (S) was prepared using starch as the fuel and LiMn 2 O 4 (P) was prepared using poly vinyl alcohol (PVA) as the fuel. XRD studies indicated a significant and consistent shift in the 2θ values of all the hkl peaks to higher values in LiMn 2 O 4 (P) compared to LiMn 2 O 4 (S) indicating a lattice contraction in the former. TG/DTA studies indicated a higher formation temperature (∼25 deg. C higher) for LiMn 2 O 4 (P). The higher formation temperature most likely promotes the oxidation of some Mn 3+ to Mn 4+ with a lower ionic radius causing a lattice contraction. This hypothesis is confirmed through XPS studies which indicated the presence of a higher fraction of Mn 4+ in LiMn 2 O 4 (P) than that present in LiMn 2 O 4 (S). A crystal shape algorithm was used to generate the crystal habits of lithium manganate from their XRD data leading to an understanding on the exposed hkl planes in these materials. From the atomic arrangement on the exposed hkl planes it is predicted that LiMn 2 O 4 (P) would be less prone to manganese dissolution and hence would possess a higher cycle life when compared to LiMn 2 O 4 (S).

  8. On the observation of a huge lattice contraction and crystal habit modifications in LiMn 2O 4 prepared by a fuel assisted solution combustion

    Science.gov (United States)

    Ragavendran, K.; Sherwood, D.; Vasudevan, D.; Emmanuel, Bosco

    2009-08-01

    Two batches of poly-crystalline lithium manganate were prepared by a fuel assisted solution combustion method. LiMn 2O 4(S) was prepared using starch as the fuel and LiMn 2O 4(P) was prepared using poly vinyl alcohol (PVA) as the fuel. XRD studies indicated a significant and consistent shift in the 2 θ values of all the hkl peaks to higher values in LiMn 2O 4(P) compared to LiMn 2O 4(S) indicating a lattice contraction in the former. TG/DTA studies indicated a higher formation temperature (∼25 °C higher) for LiMn 2O 4(P). The higher formation temperature most likely promotes the oxidation of some Mn 3+ to Mn 4+ with a lower ionic radius causing a lattice contraction. This hypothesis is confirmed through XPS studies which indicated the presence of a higher fraction of Mn 4+ in LiMn 2O 4(P) than that present in LiMn 2O 4(S). A crystal shape algorithm was used to generate the crystal habits of lithium manganate from their XRD data leading to an understanding on the exposed hkl planes in these materials. From the atomic arrangement on the exposed hkl planes it is predicted that LiMn 2O 4(P) would be less prone to manganese dissolution and hence would possess a higher cycle life when compared to LiMn 2O 4(S).

  9. How can fossil fuel based public bus transport systems become a sustainable solution for Swedish medium-sized cities?

    OpenAIRE

    Borén, Sven; Nurhadi, Lisiana; Ny, Henrik

    2013-01-01

    Vehicles, infrastructure, fuel systems and other energy-driven systems that serve public transport are complex with many resource inputs and outputs, and involve many processes. Life Cycle Assessment (LCA) and Life Cycle Costing (LCC) helps analyzing those by quantifying environmental and economic effects, but will not in themselves provide a full systems perspective. Swedish authorities have set ambitious national goals, and many regions targets a 100% increase in public transport by 2020. T...

  10. Aircraft and Bases Powered by Compact Nuclear Reactors: Solutions to Projecting Power in Highly Contested Environments and Fossil Fuel Dependence

    Science.gov (United States)

    2015-05-01

    decline.20 Since 2008, improvements in fossil fuel extraction techniques, such as fracking in the United States, have delayed the inevitable and probably...higher US production due to fracking and inaction by the Organization of the Petroleum Exporting Countries (OPEC).25 Assuming this relaxation does not... fracking techniques, eventually even the most ingenious extraction techniques will not be enough for supply to keep up with demand, and humans will

  11. Spray-on polyvinyl alcohol separators and impact on power production in air-cathode microbial fuel cells with different solution conductivities

    KAUST Repository

    Hoskins, Daniel L.

    2014-11-01

    © 2014 Elsevier Ltd. Separators are used to protect cathodes from biofouling and to avoid electrode short-circuiting, but they can adversely affect microbial fuel cell (MFC) performance. A spray method was used to apply a polyvinyl alcohol (PVA) separator to the cathode. Power densities were unaffected by the PVA separator (339 ± 29 mW/m2), compared to a control lacking a separator in a low conductivity solution (1mS/cm) similar to wastewater. Power was reduced with separators in solutions typical of laboratory tests (7-13 mS/cm), compared to separatorless controls. The PVA separator produced more power in a separator assembly (SEA) configuration (444 ± 8 mW/m2) in the 1mS/cm solution, but power was reduced if a PVA or wipe separator was used in higher conductivity solutions with either Pt or activated carbon catalysts. Spray and cast PVA separators performed similarly, but the spray method is preferred as it was easier to apply and use.

  12. The impact of new cathode materials relative to baseline performance of microbial fuel cells all with the same architecture and solution chemistry

    KAUST Repository

    Yang, Wulin

    2017-04-21

    Differences in microbial fuel cell (MFC) architectures, materials, and solution chemistries, have previously hindered direct comparisons of improvements in power production due to new cathode materials. However, one common reactor design has now been used in many different laboratories around the world under similar operating conditions based on using: a graphite fiber brush anode, a platinum cathode catalyst, a single-chamber cube-shaped (4-cm) MFC with a 3-cm diameter anolyte chamber, 50 mM phosphate buffer, and an acetate fuel. Analysis of several publications over 10 years from a single laboratory showed that even under such identical operational conditions, maximum power densities varied by 15%, with an average of 1.36 ± 0.20 W m–2 (n=24), normalized to cathode projected area (34 W m–3 liquid volume). In other laboratories, maximum power was significantly less, with an average of 1.03 ± 0.46 W m–2 (n=11), despite identical conditions. One likely reason for the differences in power is cathode age. Power production with Pt catalyst cathodes significantly declined after one month of operation or more to 0.87 ± 0.31 W m–2 (n=18) based on studies where cathode aging was examined, while in many studies the age of the cathode was not reported. Using these studies as a performance baseline, we review the claims of improvements in power generation due to new anode or cathode materials, or changes in solution conductivities and substrates.

  13. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  14. Study of Maxwell–Wagner (M–W) relaxation behavior and hysteresis observed in bismuth titanate layered structure obtained by solution combustion synthesis using dextrose as fuel

    International Nuclear Information System (INIS)

    Subohi, Oroosa; Shastri, Lokesh; Kumar, G.S.; Malik, M.M.; Kurchania, Rajnish

    2014-01-01

    Graphical abstract: X-ray diffraction studies show that phase formation and crystallinity was reached only after calcinations at 800 °C. Dielectric constant versus temperature curve shows ferroelectric to paraelectric transition temperature (T c ) to be 650 °C. Complex impedance curves show deviation from Debye behavior. The material shows a thin PE Loop with low remnant polarization due to high conductivity in the as prepared sample. - Highlights: • Bi 4 Ti 3 O 12 is synthesized using solution combustion technique with dextrose as fuel. • Dextrose has high reducing capacity (+24) and generates more no. of moles of gases. • Impedance studies show that the sample follows Maxwell–Wagner relaxation behavior. • Shows lower remnant polarization due to higher c-axis ratio. - Abstract: Structural, dielectric and ferroelectric properties of bismuth titanate (Bi 4 Ti 3 O 12 ) obtained by solution combustion technique using dextrose as fuel is studied extensively in this paper. Dextrose is used as fuel as it has high reducing valancy and generates more number of moles of gases during the reaction. X-ray diffraction studies show that phase formation and crystallinity was reached only after calcinations at 800 °C. Dielectric constant versus temperature curve shows ferroelectric to paraelectric transition temperature (T c ) to be 650 °C. The dielectric loss is very less (tan δ < 1) at lower temperatures but increases around T c due to structural changes in the sample. Complex impedance curves show deviation from Debye behavior. The material shows a thin PE Loop with low remnant polarization due to high conductivity in the as prepared sample

  15. Dissolution performance of plutonium nitride based fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Aneheim, E.; Hedberg, M. [Nuclear Chemistry, Chemistry and Chemical Engineering, Chalmers University of Technology, Kemivaegen 4, Gothenburg, SE41296 (Sweden)

    2016-07-01

    Nitride fuels have been regarded as one viable fuel option for Generation IV reactors due to their positive features compared to oxides. To be able to close the fuel cycle and follow the Generation IV concept, nitrides must, however, demonstrate their ability to be reprocessed. This means that the dissolution performance of actinide based nitrides has to be thoroughly investigated and assessed. As the zirconium stabilized nitrides show even better potential as fuel material than does the pure actinide containing nitrides, investigations on the dissolution behavior of both PuN and (Pu,Zr)N has been undertaken. If possible it is desirable to perform the fuel dissolutions using nitric acid. This, as most reprocessing strategies using solvent-solvent extraction are based on a nitride containing aqueous matrix. (Pu,Zr)N/C microspheres were produced using internal gelation. The spheres dissolution performance was investigated using nitric acid with and without additions of HF and Ag(II). In addition PuN fuel pellets were produced from powder and their dissolution performance were also assessed in a nitric acid based setting. It appears that both PuN and (Pu,Zr)N/C fuel material can be completely dissolved in nitric acid of high concentration with the use of catalytic amounts of HF. The amount of HF added strongly affects dissolution kinetics of (Pu, Zr)N and the presence of HF affects the 2 solutes differently, possibly due to inhomogeneity o the initial material. Large additions of Ag(II) can also be used to facilitate the dissolution of (Pu,Zr)N in nitric acid. PuN can be dissolved by pure nitric acid of high concentration at room temperature while (Pu, Zr)N is unaffected under similar conditions. At elevated temperature (reflux), (Pu,Zr)N can, however, also be dissolved by concentrated pure nitric acid.

  16. Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane; Eapen, Jacob

    2013-10-01

    Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt

  17. Process for dissolving the radioactive corrosion products from internal surfaces in nuclear reactors

    International Nuclear Information System (INIS)

    Brown, W.W.

    1976-01-01

    This invention concerns a process for dissolving in the coolant flowing in a reactor the radioactive substances from the corrosion of the internal surfaces of the reactor to which they cling. When a reactor is operating, the fission occurring in the fuel generates gases and fission substances, such as iodine 131 and 133, cesium 134 and 137, molybdenum 99, xenon 133 and activates the structural materials of the reactor such as nickel by giving off cobalt 58 and similar substances. Under this invention an oxygen rich solution is injected in the reactor coolant after the temperature and pressure reduction stage, during the preparation prior to refuelling and repairs. The oxygen in the solution speeds up the release of cobalt 58 and other radioactive substances from the internal surfaces of the reactor and their dissolving in the oxygenated cold coolant at the start of the cooling procedures of the installation. This allows them to be removed by an ion exchanger before the reactor is emptied. By utilising this process, about half a day may be gained in refuelling time when this has to be done once a week [fr

  18. Determination of hexamethylene tetramine in the process solution of sol-gel method for nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Ganatra, V.R.; Sawant, R.M.; Chaudhuri, N.K.; Vaidya, V.N.

    1998-01-01

    Hexamethylene tetramine (HMTA) was determined in the presence of large quantities of urea, formaldehyde and ammonium hydroxide by potentiometric titration with perchloric acid solution using an autotitrator coupled to a personal computer. This analysis is required for the process control of the sol-gel method in the production of ceramic metal oxide (e.g., oxides and mixed oxides of Th, U and Pu) microspheres using the internal gelation route. Feed solution used for preparation of microspheres contains large quantities of urea. The washings of gel microspheres produced after the internal gelation process contain urea, formaldehyde, urea-formaldehyde complex and ammonium hydroxide. The presence of these constituents in the feed solution and washings seriously interfere in the commonly used methods for the determination of HMTA. Using this method the relative standard deviation was found to be 0.27% in eleven determinations of a typical feed solution (3.0M HMTA) when the aliquots contained 75 to 125 mg of HMTA. Time required for each titration was 5-7 minutes. Feed and effluent solutions of sol-gel process were analysed. (author)

  19. Pollution from Fossil-Fuel Combustion is the Leading Environmental Threat to Global Pediatric Health and Equity: Solutions Exist.

    Science.gov (United States)

    Perera, Frederica

    2017-12-23

    Fossil-fuel combustion by-products are the world's most significant threat to children's health and future and are major contributors to global inequality and environmental injustice. The emissions include a myriad of toxic air pollutants and carbon dioxide (CO₂), which is the most important human-produced climate-altering greenhouse gas. Synergies between air pollution and climate change can magnify the harm to children. Impacts include impairment of cognitive and behavioral development, respiratory illness, and other chronic diseases-all of which may be "seeded" in utero and affect health and functioning immediately and over the life course. By impairing children's health, ability to learn, and potential to contribute to society, pollution and climate change cause children to become less resilient and the communities they live in to become less equitable. The developing fetus and young child are disproportionately affected by these exposures because of their immature defense mechanisms and rapid development, especially those in low- and middle-income countries where poverty and lack of resources compound the effects. No country is spared, however: even high-income countries, especially low-income communities and communities of color within them, are experiencing impacts of fossil fuel-related pollution, climate change and resultant widening inequality and environmental injustice. Global pediatric health is at a tipping point, with catastrophic consequences in the absence of bold action. Fortunately, technologies and interventions are at hand to reduce and prevent pollution and climate change, with large economic benefits documented or predicted. All cultures and communities share a concern for the health and well-being of present and future children: this shared value provides a politically powerful lever for action. The purpose of this commentary is to briefly review the data on the health impacts of fossil-fuel pollution, highlighting the neurodevelopmental

  20. Polygeneration microgrids: A viable solution in remote areas for supplying power, potable water and hydrogen as transportation fuel

    International Nuclear Information System (INIS)

    Kyriakarakos, George; Dounis, Anastasios I.; Rozakis, Stelios; Arvanitis, Konstantinos G.; Papadakis, George

    2011-01-01

    Highlights: → Polygeneration of power, hydrogen and potable water through desalination in remote areas. → Particle Swarm Optimization for the design of Polygeneration microgrid design with TRNSYS, GenOpt and TRNOPT. → Economic evaluation with Monte Carlo simulation for the calculation of NPV distribution. → Polygeneration microgrids are technically feasible and most likely financially profitable. -- Abstract: This paper presents the concept and the design of a hybrid renewable energy polygeneration microgrid along with its technical and economical evaluation. The energy of the sun and the wind is harvested by photovoltaics and a wind turbine. Besides that, the components of the microgrid include a battery bank, a Proton Exchange Membrane (PEM) fuel cell, a PEM electrolyzer, a metal hydride tank, a reverse osmosis desalination unit using energy recovery and a control system. The microgrid covers the electricity, transport and water needs and thus its products are power, hydrogen as transportation fuel and potable water through desalination. Hydrogen and the desalinated water also act as medium to long term seasonal storage. A design tool based on TRNSYS 16, GenOpt 2.0 and TRNOPT was developed using Particle Swarm Optimization method. The economic evaluation of the concept was based on the discounting cash flow approach. The Monte Carlo Simulation method was used in order to take uncertainty into account. A technically feasible polygeneration microgrid adapted to a small island is financially profitable with a probability of 90% for the present and 100% at the medium term.

  1. Sequential determination of urea and HMTA in the process solution of sol-gel route of advanced nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Sawant, R.M.; Chaudhuri, N.K.; Ramakumar, K.L.

    2002-01-01

    Determinations of hexamethylene tetramine (HMTA) and urea in the process solutions are required to optimize their concentrations for obtaining high quality ceramic oxide microspheres, for monitoring the washing procedure and for their subsequent recovery, recycling or waste disposal. Determination of urea is the feed solution by conventional procedures is difficult as it contains HMTA. It is more so in the effluent as it contains hydrolytic products like formaldehyde, methylol derivatives of urea, ammonium nitrate and ammonium hydroxide used for washing the gel microspheres. A derivative potentiometric method using a microprocessor-based autotitrator is described. Peaks on the first derivative of the titration plot corresponded to constituents of different basicities. Urea was selectively hydrolyzed at room temperature by the catalytic action of urease enzyme leaving HMTA unaffected. Ammonium hydroxide and ammonium bicarbonate produced from urea and HMTA were sequentially titrated for the analysis of the feed solution to obtain the three corresponding peaks respectively. Two separate titrations were required for the analysis of the effluent solution, which contained free ammonia also. One aliquot was first titrated directly without adding urease (for free ammonia and HMTA) and another aliquot was titrated after treatment with urease. The end points due to the ammonia used for washing and that from urea hydrolysis merged resulting in the appearance of three peaks again. Using this sequential method the relative standard deviations were found to be 0.81% and 1.38% for urea and HMTA, respectively, in eight determinations when the aliquots contained 50 to 75 mg of urea and 75 to 125 mg of HMTA. Feed and effluent solutions of the process stream were analyzed. (author)

  2. Studies of dissolution solutions of ruthenium metal, oxide and mixed compounds in nitric acid

    International Nuclear Information System (INIS)

    Mousset, F.; Eysseric, C.; Bedioui, F.

    2004-01-01

    Ruthenium is one of the fission products generated by irradiated nuclear fuel. It is present throughout all the steps of nuclear fuel reprocessing-particularly during extraction-and requires special attention due to its complex chemistry and high βγ activity. An innovative electro-volatilization process is now being developed to take advantage of the volatility of RuO 4 in order to eliminate it at the head end of the Purex process and thus reduce the number of extraction cycles. Although the process operates successfully with synthetic nitrato-RuNO 3+ solutions, difficulties have been encountered in extrapolating it to real-like dissolution solutions. In order to better approximate the chemical forms of ruthenium found in fuel dissolution solutions, kinetic and speciation studies on dissolved species were undertaken with RuO 2 ,xH 2 O and Ru 0 in nitric acid media. (authors)

  3. Biological processes for environmental control of effluent streams in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Shumate, S.E. II; Hancher, C.W.; Strandberg, G.W.; Scott, C.D.

    1978-01-01

    Nitrates and radioactive heavy metals need to be removed from aqueous effluent streams in the fuel cycle. Biological methods are being developed for reducing nitrate or nitrite to N 2 gas and for decreasing dissolved metal concentration to less than 1 g/m 3 . Fluidized-bed denitrification bioreactors are being tested. Removal of uranium from solution by Saccharomyces cerevisiae and Pseudomonas aeruginosa was studied

  4. Flow sheet development for the dissolution of unirradiated Mark 42 fuel tubes in F-Canyon, Part II

    International Nuclear Information System (INIS)

    Murray, A.M.

    1999-01-01

    Two dissolution flow sheets were tested for the desorption of unirradiated Mark 42 fuel tubes. Both the aluminum (from the can, cladding, and fuel core) and the plutonium oxide (PuO 2 ) are dissolved simultaneously, i.e., a co-dissolution flow sheet. In the first series of tests, 0.15 and 0.20 molar (M) potassium fluoride (KF) solutions were used and the dissolution extended over several days. In the other series of tests, solutions with higher concentrations of fluoride (0.25 to 0.30 M) were used. Calcium fluoride (CaF 2 ) was used in those tests as the fluoride source

  5. Characterization of Ni-YSZ anodes for solid oxide fuel cells fabricated by solution precursor plasma spraying with axial feedstock injection

    Science.gov (United States)

    Metcalfe, Craig; Lay-Grindler, Elisa; Kesler, Olivera

    2014-02-01

    Nickel and yttria-stabilized zirconia (YSZ) anodes were fabricated by solution precursor plasma spraying (SPPS) and incorporated into metal-supported solid oxide fuel cells (SOFC). A power density of 0.45 W cm-2 at 0.7 V and a peak power density of 0.52 W cm-2 at 750 °C in humidified H2 was obtained, which are the first performance results reported for an SOFC having an anode fabricated by SPPS. The effects of solution composition, plasma gas composition, and stand-off distance on the composition of the deposited Ni-YSZ coatings by SPPS were evaluated. It was found that the addition of citric acid to the aqueous solution delayed re-solidification of NiO particles, improving the deposition efficiency and coating adhesion. The composition of the deposited coatings was found to vary with torch power. Increasing torch power led to coatings with decreasing Ni content, as a result of Ni vaporizing in-flight at stand-off distances less than 60 mm from the torch nozzle exit.

  6. Pollution from Fossil-Fuel Combustion is the Leading Environmental Threat to Global Pediatric Health and Equity: Solutions Exist

    Science.gov (United States)

    Perera, Frederica

    2017-01-01

    Fossil-fuel combustion by-products are the world’s most significant threat to children’s health and future and are major contributors to global inequality and environmental injustice. The emissions include a myriad of toxic air pollutants and carbon dioxide (CO2), which is the most important human-produced climate-altering greenhouse gas. Synergies between air pollution and climate change can magnify the harm to children. Impacts include impairment of cognitive and behavioral development, respiratory illness, and other chronic diseases—all of which may be “seeded“ in utero and affect health and functioning immediately and over the life course. By impairing children’s health, ability to learn, and potential to contribute to society, pollution and climate change cause children to become less resilient and the communities they live in to become less equitable. The developing fetus and young child are disproportionately affected by these exposures because of their immature defense mechanisms and rapid development, especially those in low- and middle-income countries where poverty and lack of resources compound the effects. No country is spared, however: even high-income countries, especially low-income communities and communities of color within them, are experiencing impacts of fossil fuel-related pollution, climate change and resultant widening inequality and environmental injustice. Global pediatric health is at a tipping point, with catastrophic consequences in the absence of bold action. Fortunately, technologies and interventions are at hand to reduce and prevent pollution and climate change, with large economic benefits documented or predicted. All cultures and communities share a concern for the health and well-being of present and future children: this shared value provides a politically powerful lever for action. The purpose of this commentary is to briefly review the data on the health impacts of fossil-fuel pollution, highlighting the

  7. Dissolution of FFTF vendor fuel

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Dissolution experiments were performed on FFTF vendor fuel (both mechanically mixed and coprecipitated) during 1974, 1975, and 1976. A marked improvement was noted in the completeness of fuel dissolution from 1974 to 1976. The reason for this is unknown but may have been attributable to slight changes in fuel fabrication conditions. In general, the bulk of the fuel pellets tested dissolved to greater than 99.9% in nitric acid alone

  8. Dissolution of FFTF vendor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lerch, R.E.

    1979-08-01

    Dissolution experiments were performed on FFTF vendor fuel (both mechanically mixed and coprecipitated) during 1974, 1975, and 1976. A marked improvement was noted in the completeness of fuel dissolution from 1974 to 1976. The reason for this is unknown but may have been attributable to slight changes in fuel fabrication conditions. In general, the bulk of the fuel pellets tested dissolved to greater than 99.9% in nitric acid alone.

  9. Dissolve energy obesity by energy diet

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Heum [Sunmoon University, Asan (Korea)

    2000-07-01

    Every organism takes needed materials or energy from outside and excretes unessential things to outside. This is called a metabolism or energy metabolism. Calculating the amount of energy consumed by human in the world by converting to the amount of metabolism of an animal to survive, the weight of a human being is corresponding to an animal with a weigh of 40 ton. Human beings can find a solution to dissolve energy obesity or can maintain a massive status by finding a new energy source in the universe.

  10. Hydrothermal synthesis for fabrication and reprocessing of MOX nuclear fuel

    International Nuclear Information System (INIS)

    Ohta, Suguru; Yamamura, Tomoo; Shirasaki, Kenji; Satoh, Isamu; Shikama, Tatsuo

    2011-01-01

    To improve the nuclear proliferation resistance and to minimize use of chemicals, a new reprocessing and fabrication process of 'mixed oxide' (MOX) fuel was proposed and studied by using simulated spent fuel solutions. The process is consisting of the two steps, i.e. the removal of fission product (FP) from dissolved spent fuel by using carbonate solutions (Step-1), and hydrothermal synthesis of uranium dioxides (Step-2). In Step-1, rare earth (the precipitation ratio: 90%) and alkaline earth (10-50% for Sr) as FP were removed based on their low solubility of hydroxides and carbonate salts, with uranium kept dissolved for the certain carbonate solutions of weak base (Type 2) or mixtures of relatively strong base and weak base (Type 3). In Step-2, the features of uranium dioxides UO 2+x particles, i.e. stoichiometry (x=0.05-0.2), size (0.2-3 μm) and shape (cubic, spherical, rectangular parallelpiped, etc.), were controlled, and the cesium was removed down to 40 ppm by an addition of organic additives. The decontamination factors (DF) for cesium exceeds 10 5 , whereas the total DF of all the simulated FP were as low as the order of 10 which requires future studies for removal of alkaline earth, Re and Tc etc. (author)

  11. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    (IV) oxyhydroxides species due to radiolytic effects cannot completely be ruled out. Solution concentrations of U were within the range of the solubility limits of the solid phase U(OH){sub 4}(am). The determined concentrations of U and Am in solution were about one order of magnitude higher for the U{sub 3}Si{sub 2}-Al fuel sample. Here, the formation of U/Si containing secondary phase components and their influence on radionuclide solubility cannot be ruled out. Results of this work show that the U{sub 3}Si{sub 2}-Al and UAl{sub x}-Al dispersed research reactor spent fuel samples dissolved completely within the test period of 3.5 years in MgCl{sub 2}-rich brine in the presence of Fe{sup 2+}. In view of final disposal this means that these fuel matrices represent no barrier. The radionuclides will be released instantaneously. Cs (the long-lived isotope {sup 135}Cs is of special concern with respect to final disposal) and Sr were classified as mobile radionuclide species. For U, Am, Pu and Eu, a reimmobilization was observed. Sorption is the process which is assumed to be responsible for the reimmobilization of the long-lived actinide Am and the lanthanide Eu. Solution concentrations of U and Pu seem to be controlled by their solubility controlling solid phases.

  12. Effect of fuels on conductivity, dielectric and humidity sensing properties of ZrO2 nanocrystals prepared by low temperature solution combustion method

    Directory of Open Access Journals (Sweden)

    H.C. Madhusudhana

    2016-09-01

    Full Text Available ZrO2 nanopowders were synthesized by low temperature solution combustion method using two different fuels namely glycine and oxalyldihydrazide (ODH. The phase confirmation was done by powder X-ray diffraction (PXRD and Raman spectral analysis. Use of glycine resulted in ZrO2 with mixture of tetragonal and monoclinic phase with average crystallite size of ∼30 nm. However, ODH as fuel aids in the formation of ZrO2 with mixture of tetragonal and cubic phase with average crystallite size ∼20 nm. Further, in present work we present novel way to tune conductivity property of the nano ZrO2. We show that merely changing the fuel from glycine to ODH, we obtain better DC conductivity and dielectric constant. On the other hand use of glycine leads to the formation of ZrO2 with better AC conductivity and humidity sensing behavior. The dielectric constants calculated for samples prepared with glycine and ODH were found to be 45 and 26 respectively at 10 MHz. The AC and DC conductivity values of the samples prepared with glycine was found to be 9.5 × 10−4 S cm−1, 1.1 × 10−3 S cm−1 and that of ODH was 7.6 × 10−4 S cm−1, 3.6 × 10−3 S cm−1 respectively.

  13. Engineering solutions for a reflector change concept in the high-temperature reactor with pebble bed core and OTTO-fueling

    International Nuclear Information System (INIS)

    Kasper, K.J.

    1975-06-01

    In the field of reactor engineering an increasing tendency is visible towards a 'repairable reactor'. In the construction of the HTR with spherical fuel elements this fact should already be taken into account at an early stage. Additionally it is possible that in connection with the OTTO-fueling load conditions for the graphite reflector could result which are locally not far away from limiting values. Therefore the removability of the reflector is included in the reactor construction as an accompanying technical step of the physical lay-out of the core. The core arrangements, realized for HTR until recently, are discussed as well as the properties of the graphites used and the operating conditions in the reactors are stated. At the example of the PR 3,000 proposals are offered for the construction of a removable side and top reflector for a pebble bed reactor. Hereby a solution was found which, on one hand allows the changing of the reflector and on the other hand requires no significant increase of the costs for the reactor assembly. Moreover the requirements of reactor operation and of repairability are satisfied in an optimal manner. (orig.) [de

  14. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

  15. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

    1979-04-01

    Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

  16. Collective processing device for spent fuel

    International Nuclear Information System (INIS)

    Irie, Hiroaki; Taniguchi, Noboru.

    1996-01-01

    The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

  17. METHOD OF DISSOLVING URANIUM METAL

    Science.gov (United States)

    Slotin, L.A.

    1958-02-18

    This patent relates to an economicai means of dissolving metallic uranium. It has been found that the addition of a small amount of perchloric acid to the concentrated nitric acid in which the uranium is being dissolved greatly shortens the time necessary for dissolution of the metal. Thus the use of about 1 or 2 percent of perchioric acid based on the weight of the nitric acid used, reduces the time of dissolution of uranium by a factor of about 100.

  18. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    International Nuclear Information System (INIS)

    Rudisill, T; John Mickalonis, J

    2006-01-01

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

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

    International Nuclear Information System (INIS)

    Ollila, K.

    1995-11-01

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

  20. Iodine and NOx behavior in the dissolver off-gas and IODOX [Iodine Oxidation] systems in the Oak Ridge National Laboratory Integrated Equipment Test facility

    International Nuclear Information System (INIS)

    Birdwell, J.F.

    1990-01-01

    This paper describes the most recent in a series of experiments evaluating the behavior of iodine and NO x in the Integrated Equipment Test (IET) Dissolver Off-Gas (DOG) System. This work was performed as part of a joint collaborative program between the US Department of Energy and the Power and Nuclear Fuel Development Corporation of Japan. The DOG system consists of two shell-and-tube heat exchangers in which water and nitric acid are removed from the dissolver off-gas by condensation, followed by a packed tower in which NO x is removed by absorption into a dilute nitric acid solution. The paper also describes the results of the operation of the Iodine Oxidation (IODOX) System. This system serves to remove iodine from the DOG system effluent by absorption into hyperazeotropic nitric acid. 7 refs., 11 figs., 10 tabs

  1. Handling and storage of high-level liquid wastes from reprocessing of spent fuel

    International Nuclear Information System (INIS)

    Finsterwalder, L.

    1982-01-01

    The high level liquid wastes arise from the reprocessing of irradiated nuclear fuels, which are dissolved in aqueous acid solution, and the plutonium and unburned uranium removed in the chemical separation plant. The remaining solution, containing more than 99% of the dissolved fission products, together with impurities from cladding materials, corrosion products, traces of unseparated plutonium and uranium and most of the transuranic elements, constitutes the high-level waste. At present, these liquid wastes are usually concentrated by evaporation and stored as an aqueous nitric acid solution in high-integrity stainless-steel tanks. There is now world-wide agreement that, for the long term, these liquid wastes should be converted to solid form and much work is in progress to develop techniques for the solidification of these wastes. This paper considers the design requirements for such facilities and the experience gained during nearly 30 years of operation. (orig./RW)

  2. Preparation of encapsulated proteins dissolved in low viscosity fluids

    International Nuclear Information System (INIS)

    Ehrhardt, Mark R.; Flynn, Peter F.; Wand, A. Joshua

    1999-01-01

    The majority of proteins are too large to be comprehensively examined by solution NMR methods, primarily because they tumble too slowly in solution. One potential approach to making the NMR relaxation properties of large proteins amenable to modern solution NMR techniques is to encapsulate them in a reverse micelle which is dissolved in a low viscosity fluid. Unfortunately, promising low viscosity fluids such as the short chain alkanes, supercritical carbon dioxide, and various halocarbon refrigerants all require the application of significant pressure to be kept liquefied at room temperature. Here we describe the design and use of a simple cost effective NMR tube suitable for the preparation of solutions of proteins encapsulated in reverse micelles dissolved in such fluids

  3. Isotope correlation verification of analytical measurements for dissolver materials

    International Nuclear Information System (INIS)

    Satkowski, J.

    1988-01-01

    An independent verification of analytical results for accountability measurements of dissolver materials can be performed using the Iosotop Correlation Technique (ICT). ICT is based on the relationships that exist between the initial and final elemental concentration and isotopic abundances of the nuclear fuel. Linear correlation functions between isotopic ratios and plutonium/uranium ratios have been developed for specific reactor fuels. The application of these correlations to already existing analytical data provides a laboratory additional confidence in the reported results. Confirmation is done by a test of consistancy with historical data. ICT is being utilized with dissolver accountability measurements at the Savannah River Plant Laboratory. The application, implementation, and operating experience of this technique are presented

  4. Balance and behavior of gaseous radionuclides released during initial PWR fuel reprocessing operations

    International Nuclear Information System (INIS)

    Leudet, A.; Miquel, P.; Goumondy, P.J.; Charrier, G.

    1982-08-01

    Five fuel pins, taken from a PWR fuel assembly with 32000 MWD/t burn-up were chopped and dissolved in leak-proof equipment designed for accurate determination of the composition and quantity of gaseous elements released in these operations. Analytical methods were specially developped to determine directly the noble gases, tritium and gaseous carbon compounds in the gas phase. Volatile iodine was kept as close as possible to the source by cold traps, then transferred to a caustic solution for quantitative analysis. The quantities and activities of gaseous fission products thus determined were compared with predicted values obtained through computation. Very good agreement was generally observed

  5. Balance and behavior of gaseous radionuclides released during initial PWR fuel reprocessing operations

    International Nuclear Information System (INIS)

    Leudet, A.; Miquel, P.; Goumondy, P.J.; Charrier, G.

    1983-01-01

    Five fuel pins, taken from a PWR fuel assembly with 32,000 MwD/t burn-up were chopped and dissolved in leak-proof equipment designed for accurate determination of the composition and quantity of gaseous elements released in these operations. Analytical methods were specially developed to determine directly the noble gases, tritium and gaseous carbon compounds in the gas phase. Volatile iodine was kept as close as possible to the source by cold traps, then transferred to a caustic solution for quantitative analysis. The quantities and activities of gaseous fission products thus determined were compared with predicted values obtained through computation. Very good agreement was generally observed

  6. Irradiated uranium reprocessing, Final report I-VI, IV Deo IV - Separation of uranium, plutonium and fission products from the irradiated fuel of the reactor in Vinca; Prerada ozracenog urana. Zavrsni izvestaj - I-VI, IV Deo - Odvajanje urana, plutonijuma i fisionih produkata iz isluzenog goriva reaktora u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This study describes the technology for separation of uranium, plutonium and fission products from the radioactive water solution which is obtained by dissolving the spent uranium fuel from the reactor in Vinca. The procedure should be completed in a hot cell, with the maximum permitted activity of 10 Ci.

  7. Reducing emissions from uranium dissolving

    International Nuclear Information System (INIS)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO x emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO x fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO x emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO 2 which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered

  8. Electrochemical performance of solid oxide fuel cells having electrolytes made by suspension and solution precursor plasma spraying

    Science.gov (United States)

    Marr, M.; Kuhn, J.; Metcalfe, C.; Harris, J.; Kesler, O.

    2014-01-01

    Yttria-stabilized zirconia (YSZ) electrolytes were deposited by suspension plasma spraying (SPS) and solution precursor plasma spraying (SPPS). The electrolytes were evaluated for permeability, microstructure, and electrochemical performance. With SPS, three different suspensions were tested to explore the influence of powder size distribution and liquid properties. Electrolytes made from suspensions of a powder with d50 = 2.6 μm were more gas-tight than those made from suspensions of a powder with d50 = 0.6 μm. A peak open circuit voltage of 1.00 V was measured at 750 °C with a cell with an electrolyte made from a suspension of d50 = 2.6 μm powder. The use of a flammable suspension liquid was beneficial for improving electrolyte conductivity when using lower energy plasmas, but the choice of liquid was less important when using higher energy plasmas. With SPPS, peak electrolyte conductivities were comparable to the peak conductivities of the SPS electrolytes. However, leak rates through the SPPS electrolytes were higher than those through the electrolytes made from suspensions of d50 = 2.6 μm powder. The electrochemical test data on SPPS electrolytes are the first reported in the literature.

  9. Source Reduction Effectiveness at Fuel Contaminated Sites, Technical Summary Report

    National Research Council Canada - National Science Library

    2000-01-01

    This report assesses the degree to which various types or engineered source-reduction efforts at selected fuel-contaminated sites have resulted in decreasing concentrations of fuel constituents dissolved in groundwater...

  10. Nuclear fuel cycle studies

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    For the metal-matrix encapsulation of radioactive waste, brittle-fracture, leach-rate, and migration studies are being conducted. For fuel reprocessing, annular and centrifugal contactors are being tested and modeled. For the LWBR proof-of-breeding project, the full-scale shear and the prototype dissolver were procured and tested. 5 figures

  11. 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.

  12. CADDIS Volume 2. Sources, Stressors and Responses: Dissolved Oxygen

    Science.gov (United States)

    Introduction to the dissolved oxygen module, when to list dissolved oxygen as a candidate cause, ways to measure dissolved oxygen, simple and detailed conceptual model diagrams for dissolved oxygen, references for the dissolved oxygen module.

  13. Transuranium element recovering method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Todokoro, Akio; Kihara, Yoshiyuki; Okada, Hisashi

    1998-01-01

    Spent fuels are dissolved in nitric acid, the obtained dissolution liquid is oxidized by electrolysis, and nitric acid of transuranium elements are precipitated together with nitric acid of uranium elements from the dissolution solution and recovered. Namely, the transuranium elements are oxidized to an atomic value level at which nitric acid can be precipitated by an oxidizing catalyst, and cooled to precipitate nitric acid of transuranium elements together with nitric acid of transuranium elements, accordingly, it is not necessary to use a solvent which has been used so far upon recovering transuranium elements. Since no solvent waste is generated, a recovery method taking the circumstance into consideration can be provided. Further, nitric acid of uranium elements and nitric acid of transuranium elements precipitated and recovered together are dissolved in nitric acid again, cooled and only uranium elements are precipitated selectively, and recovered by filtration. The amount of wastes can be reduced to thereby enabling to mitigate control for processing. (N.H.)

  14. Achievement report on research and development in the Sunshine Project in fiscal 1979. Research on fuel cells (Research on aqueous alkaline solution electrolyte fuel cells); 1979 nendo nenryo denchi no kenkyu seika hokokusho. Arukari suiyoeki denkaishitsu nenryo denchi no kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-03-01

    This paper describes achievements in fiscal 1979 in research on aqueous alkaline solution electrolyte fuel cells. Trial fabrication and tests for an oxygen electrode were performed on a catalytic electrode added with silver using carbonblack and graphites as carriers having excellent corrosion resistance and large surface area. Characteristics not inferior to electrodes using activated carbon as a carrier were obtained in both of the initial characteristics and continuous discharge characteristics. A platinum added electrode also showed the same performance as the silver added electrode. A hydrogen electrode containing Zr and iron among those containing Raney-Ni was found to have high oxidation resistance and stability in terms of life. A platinum added electrode using graphite as a carrier provided satisfactory initial characteristics as a hydrogen electrode. Research on a single cell construction has used and tested eight-cell laminated cells with an area of 1,000 cm{sup 2} using bipolar sheets made of carbon. The test verified appropriate the removal of produced water and heat using mainly the hydrogen circulation, which has been discussed in the summary design. The paper describes heat cycles, for which tests of ten and odds times in total were performed to have demonstrated that they are free of any anomaly. Furthermore, a manifold was attached as a means to improve the volume efficiency. Its function was also tested. (NEDO)

  15. Achievement report on research and development in the Sunshine Project in fiscal 1980. Research on fuel cells (Research on aqueous alkaline solution electrolyte fuel cells); 1980 nendo nenryo denchi no kenkyu seika hokokusho. Arukari suiyoeki denkaishitsu nenryo denchi no kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1981-03-01

    This paper describes the achievements in fiscal 1980 in the Sunshine Project on developing aqueous alkaline solution electrolyte fuel cells. The oxygen electrode is a silver catalyst electrode using as carriers carbon lack and graphite powders having good corrosion resistance. A continuous discharge test was performed on the electrode for 3,000 hours. Furnace black having large surface area and naturally colloidal graphite showed stable performance. The hydrogen electrode, which is a catalyst electrode made of Raney-Ni containing third elements was given a 3000-hour continuous discharge test, where Zr addition presented stable performance. Activated carbon was found a good carrier in platinum added electrodes. For electrodes operating under high pressure gases, electrodes attached with sintered nickel film are suitable. With regard to prototype cells, laminated cells with high volume efficiency structured mainly with gasket seals were designed and fabricated. The IV characteristics measurement and continuous discharge test thereon revealed no functional problems. Furthermore, cells having electrode area of 100 cm{sup 2} and using bipolar sheets made of carbon were verified capable of withstanding 2000-hour continuous discharge. A non-conductive resin mold was proposed to prevent electrolytic corrosion of the carbon sheets. Discussions were given also on corrosion resistance of bonding agents. (NEDO)

  16. Improvement of activated carbons as oxygen reduction catalysts in neutral solutions by ammonia gas treatment and their performance in microbial fuel cells

    KAUST Repository

    Watson, Valerie J.

    2013-11-01

    Commercially available activated carbon (AC) powders from different precursor materials (peat, coconut shell, coal, and hardwood) were treated with ammonia gas at 700 C to improve their performance as oxygen reduction catalysts in neutral pH solutions used in microbial fuel cells (MFCs). The ammonia treated ACs exhibited better catalytic performance in rotating ring-disk electrode tests than their untreated precursors, with the bituminous based AC most improved, with an onset potential of Eonset = 0.12 V (untreated, Eonset = 0.08 V) and n = 3.9 electrons transferred in oxygen reduction (untreated, n = 3.6), and the hardwood based AC (treated, E onset = 0.03 V, n = 3.3; untreated, Eonset = -0.04 V, n = 3.0). Ammonia treatment decreased oxygen content by 29-58%, increased nitrogen content to 1.8 atomic %, and increased the basicity of the bituminous, peat, and hardwood ACs. The treated coal based AC cathodes had higher maximum power densities in MFCs (2450 ± 40 mW m-2) than the other AC cathodes or a Pt/C cathode (2100 ± 1 mW m-2). These results show that reduced oxygen abundance and increased nitrogen functionalities on the AC surface can increase catalytic performance for oxygen reduction in neutral media. © 2013 Elsevier B.V. All rights reserved.

  17. Solution-chemical route to generalized synthesis of metal germanate nanowires with room-temperature, light-driven hydrogenation activity of CO2 into renewable hydrocarbon fuels.

    Science.gov (United States)

    Liu, Qi; Zhou, Yong; Tu, Wenguang; Yan, Shicheng; Zou, Zhigang

    2014-01-06

    A facile solution-chemical route was developed for the generalized preparation of a family of highly uniform metal germanate nanowires on a large scale. This route is based on the use of hydrazine monohydrate/H2O as a mixed solvent under solvothermal conditions. Hydrazine has multiple effects on the generation of the nanowires: as an alkali solvent, a coordination agent, and crystal anisotropic growth director. Different-percentage cobalt-doped Cd2Ge2O6 nanowires were also successfully obtained through the addition of Co(OAc)2·4H2O to the initial reaction mixture for future investigation of the magnetic properties of these nanowires. The considerably negative conduction band level of the Cd2Ge2O6 nanowire offers a high driving force for photogenerated electron transfer to CO2 under UV-vis illumination, which facilitates CO2 photocatalytic reduction to a renewable hydrocarbon fuel in the presence of water vapor at room temperature.

  18. Development of some operations in technological flowsheet for spent VVER fuel reprocessing at a pilot plant

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Galkin, B.Ya; Lyubtsev, R.I.; Romanovskii, V.N.; Velikhov, E.P.

    1981-01-01

    The fuel reprocessing pilot plants for high active materials would permit the study and development or particular processing steps and flowsheet variations; in some cases, these experimental installations realize on a small scale practically all technological chains of large reprocessing plants. Such a fuel reprocessing pilot plant with capacity of 3 kg U/d has been built at V. G. Khlopin Radium Institute. The pilot plant is installed in the hot cell of radiochemical compartment, and is composed of the equipments for fuel element cutting and dissolving, the preparation of feed solution (clarification, correction), extraction reprocessing and the production of uranium, plutonium and neptunium concentrates, the complex processing of liquid and solid wastes and a special unit for gas purification and analysis. In the last few years, a series of experiments have been carried out on the reprocessing of spent VVER fuel. (J.P.N.)

  19. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-09-01

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  20. Research on Dynamic Dissolving Model and Experiment for Rock Salt under Different Flow Conditions

    Directory of Open Access Journals (Sweden)

    Xinrong Liu

    2015-01-01

    Full Text Available Utilizing deep rock salt cavern is not only a widely recognized energy reserve method but also a key development direction for implementing the energy strategic reserve plan. And rock salt cavern adopts solution mining techniques to realize building cavity. In view of this, the paper, based on the dissolving properties of rock salt, being simplified and hypothesized the dynamic dissolving process of rock salt, combined conditions between dissolution effect and seepage effect in establishing dynamic dissolving models of rock salt under different flow quantities. Devices were also designed to test the dynamic dissolving process for rock salt samples under different flow quantities and then utilized the finite-difference method to find the numerical solution of the dynamic dissolving model. The artificial intelligence algorithm, Particle Swarm Optimization algorithm (PSO, was finally introduced to conduct inverse analysis of parameters on the established model, whose calculation results coincide with the experimental data.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jerden, James L., E-mail: jerden@anl.gov [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Frey, Kurt [University of Notre Dame, Notre Dame, IN 46556 (United States); Ebert, William [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States)

    2015-07-15

    Highlights: • This model accounts for chemistry, temperature, radiolysis, U(VI) minerals, and hydrogen effect. • The hydrogen effect dominates processes determining spent fuel dissolution rate. • The hydrogen effect protects uranium oxide spent fuel from oxidative dissolution. - Abstract: The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO{sub 2} and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO{sub 2} and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO{sub 2} and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H{sub 2}O{sub 2} and O{sub 2}). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit

  2. Effects of dissolved organic matter from a eutrophic lake on the freely dissolved concentrations of emerging organic contaminants.

    Science.gov (United States)

    Xiao, Yi-Hua; Huang, Qing-Hui; Vähätalo, Anssi V; Li, Fei-Peng; Chen, Ling

    2014-08-01

    The authors studied the effects of dissolved organic matter (DOM) on the bioavailability of bisphenol A (BPA) and chloramphenicol by measuring the freely dissolved concentrations of the contaminants in solutions containing DOM that had been isolated from a mesocosm in a eutrophic lake. The abundance and aromaticity of the chromophoric DOM increased over the 25-d mesocosm experiment. The BPA freely dissolved concentration was 72.3% lower and the chloramphenicol freely dissolved concentration was 56.2% lower using DOM collected on day 25 than using DOM collected on day 1 of the mesocosm experiment. The freely dissolved concentrations negatively correlated with the ultraviolent absorption coefficient at 254 nm and positively correlated with the spectral slope of chromophoric DOM, suggesting that the bioavailability of these emerging organic contaminants depends on the characteristics of the DOM present. The DOM-water partition coefficients (log KOC ) for the emerging organic contaminants positively correlated with the aromaticity of the DOM, measured as humic acid-like fluorescent components C1 (excitation/emission=250[313]/412 nm) and C2 (excitation/emission=268[379]/456 nm). The authors conclude that the bioavailability of emerging organic contaminants in eutrophic lakes can be affected by changes in the DOM. © 2014 SETAC.

  3. Simulation of uranium aluminide dissolution in a continuous aluminum dissolver system

    International Nuclear Information System (INIS)

    Evans, D.R.; Farman, R.F.; Christian, J.D.

    1990-01-01

    This paper reports on the Idaho Chemical Processing Plant (ICPP) which recovers highly-enriched uranium (uranium that contains at least 20 atom percent 235 U) from spent nuclear reactor fuel by dissolution of the fuel elements and extraction of the uranium from the aqueous dissolver product. Because the uranium is highly-enriched, consideration must be given to whether a critical mass can form at any stage of the process. In particular, suspended 235 U-containing particles are of special concern, due to their high density (6.8 g/cm 3 ) and due to the fact that they can settle into geometrically unfavorable configurations when not adequately mixed. A portion of the spent fuel is aluminum-alloy-clad uranium aluminide (UAl 3 ) particles, which dissolve more slowly than the cladding. As the aluminum alloy cladding dissolves in mercury-catalyzed nitric acid, UAl 3 is released. Under standard operating conditions, the UAl 3 dissolves rapidly enough to preclude the possibility of forming a critical mass anywhere in the system. However, postulated worst-case abnormal operating conditions retard uranium aluminide dissolution, and thus require evaluation. To establish safety limits for operating parameters, a computerized simulation model of uranium aluminide dissolution in the aluminum fuel dissolver system was developed

  4. Fuel management and economics

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G

    1972-11-01

    From international conference on nuclear solutions to world energy problems; Washington, District of Columbia, USA (12 Nov The low cost of the fuel cycle is the most attractive feature of the fast neutron breeder reactor. In order to achieve it a good fuel management is essential, with well balanced fixed investment and renewal fuel costs. In addition the designer can optimize the power station as a whole (fuel cycle and thermal characteristics). (auth)

  5. Natural gas and bio methane in the future fuel mix. Need of action and solution approaches for an accelerated etablishment in the traffic; Erdgas und Biomethan im kuenftigen Kraftstoffmix. Handlungsbedarf und Loesungsansaetze fuer eine beschleunigte Etablierung im Verkehr

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-01-15

    The contribution under consideration reports on the need of action and on solution attempts for an accelerated establishment of natural gas and bio methane in the future fuel mix. The authors come to the following conclusions: The energy situation and climatic situation require a stronger diversification of fuels and drives. The targets for the amount of natural gas and bio methane as a fuel are not reached yet. The characteristics of natural gas speak for an accelerated establishment in the traffic sector. The admixture of bio methane can increase the climatic, environmental and resources advantages. In order to penetrate the market all participants involved must commit themselves to a concrete 'roadmap'. The contribution shows which measures must be converted by the participants involved in order to be able to utilize fully the potentials of the employment of natural gas and bio methane in the traffic sector.

  6. Implicit analysis of the transient water flow with dissolved air

    Directory of Open Access Journals (Sweden)

    J. Twyman

    2018-01-01

    Full Text Available The implicit finite-difference method (IFDM for solving a system that transports water with dissolved air using a fixed (or variable rectangular space-time mesh defined by the specified time step method is applied. The air content in the fluid modifies both the wave speed and the Courant number, which makes it inconvenient to apply the traditional Method of Characteristics (MOC and other explicit schemes due to their impossibility to simulate the changes in magnitude, shape and frequency of the pressures train. The conclusion is that the IFDM delivers an accurate and stable solution, with a good adjustment level with respect to a classical case reported in the literature, being a valid alternative for the transient solution in systems that transport water with dissolved air.

  7. Reprocessing fuel from the Southwest Experimental Fast Oxide Reactor at the Savannah River Plant

    International Nuclear Information System (INIS)

    Gray, L.W.; Campbell, T.G.

    1985-11-01

    The irradiated fuel, reject fuel tubes, and fuel fabrication scrap from the Southwest Experimental Fast Oxide Reactor (SEFOR) were transferred to the Savannah River Plant (SRP) for uranium and plutonium recovery. The unirradiated material was declad and dissolved at SRP; dissolution was accomplished in concentrated nitric acid without the addition of fluoride. The irradiated fuel was declad at Atomics International and repacked in aluminum. The fuel and aluminum cans were dissolved at SRP using nitric acid catalyzed by mercuric nitrate. As this fuel was dissolved in nongeometrically favorable tanks, boron was used as a soluble neutron poison

  8. Assessing the Role of Dissolved Organic Phosphate on Rates of Microbial Phosphorus Cycling

    Science.gov (United States)

    Gonzalez, A. C.; Popendorf, K. J.; Duhamel, S.

    2016-02-01

    Phosphorus (P) is an element crucial to life, and it is limiting in many parts of the ocean. In oligotrophic environments, the dissolved P pool is cycled rapidly through the activity of microbes, with turnover times of several hours or less. The overarching aim of this study was to assess the flux of P from picoplankton to the dissolved pool and the role this plays in fueling rapid P cycling. To determine if specific microbial groups are responsible for significant return of P to the dissolved pool during cell lifetime, we compared the rate of cellular P turnover (cell-Pτ, the rate of cellular P uptake divided by cellular P content) to the rate of cellular biomass turnover (cellτ). High rates of P return to the dissolved pool during cell lifetime (high cell-Pτ/cellτ) indicate significant P regeneration, fueling more rapid turnover of the dissolved P pool. We hypothesized that cell-Pτ/cellτ varies widely across picoplankton groups. One factor influencing this variation may be each microbial group's relative uptake of dissolved organic phosphorus (DOP) versus dissolved inorganic phosphorus (DIP). As extracellular hydrolysis is necessary for P incorporation from DOP, this process may return more P to the dissolved pool than DIP incorporation. This leads to the question: does a picoplankton's relative uptake of DOP (versus DIP) affect the rate at which it returns phosphorus to the dissolved pool? To address this question, we compared the rate of cellular P turnover based on uptake of DOP and uptake DIP using cultured representatives of three environmentally significant picoplankton groups: Prochlorococcus, Synechococcus, and heterotrophic bacteria. These different picoplankton groups are known to take up different ratios of DOP to DIP, and may in turn make significantly different contributions to the regeneration and cycling phosphorus. These findings have implications towards our understanding of the timeframes of biogeochemical cycling of phosphorus in the

  9. Molybdenum solubility in aluminium nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    Heres, X.; Sans, D.; Bertrand, M.; Eysseric, C. [CEA, Centre de Marcoule, Nuclear Energy Division, DRCP, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Brackx, E.; Domenger, R.; Excoffier, E. [CEA, Centre de Marcoule, Nuclear Energy Division, DTEC, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Valery, J.F. [AREVA-NC, DOR/RDP, Paris - La Defense (France)

    2016-07-01

    For over 60 years, research reactors (RR or RTR for research testing reactors) have been used as neutron sources for research, radioisotope production ({sup 99}Mo/{sup 99m}Tc), nuclear medicine, materials characterization, etc... Currently, over 240 of these reactors are in operation in 56 countries. They are simpler than power reactors and operate at lower temperature (cooled to below 100 C. degrees). The fuel assemblies are typically plates or cylinders of uranium alloy and aluminium (U-Al) coated with pure aluminium. These fuels can be processed in AREVA La Hague plant after batch dissolution in concentrated nitric acid and mixing with UOX fuel streams. The aim of this study is to accurately measure the solubility of molybdenum in nitric acid solution containing high concentrations of aluminium. The higher the molybdenum solubility is, the more flexible reprocessing operations are, especially when the spent fuels contain high amounts of molybdenum. To be most representative of the dissolution process, uranium-molybdenum alloy and molybdenum metal powder were dissolved in solutions of aluminium nitrate at the nominal dissolution temperature. The experiments showed complete dissolution of metallic elements after 30 minutes long stirring, even if molybdenum metal was added in excess. After an induction period, a slow precipitation of molybdic acid occurs for about 15 hours. The data obtained show the molybdenum solubility decreases with increasing aluminium concentration. The solubility law follows an exponential relation around 40 g/L of aluminium with a high determination coefficient. Molybdenum solubility is not impacted by the presence of gadolinium, or by an increasing concentration of uranium. (authors)

  10. ASP project. Dissolved oxygen issues in ASP project

    Directory of Open Access Journals (Sweden)

    M.Y. Bondar

    2018-03-01

    Full Text Available The article presents the latest results of studies about the effect of dissolved oxygen on the efficiency of the ASP flooding project implemented by Salym Petroleum Development N.V.. Pilot project on experimental injection of anionic surfactant solutions, soda and polymer into the reservoir for enhanced oil recovery (ASP project has been implemented since 2016. The stability of one of the components of the ASP polymer is strongly dependent on the presence of iron, stiffness cations and dissolved oxygen in the water. Since the polymer is used for injection at two stages of the project, which are essential and the longest, at the design stage of ASP project a whole complex of polymer protection had been established against negative factors, in particular from the influence of oxygen, which causes not only oxygen corrosion but also irreversible destruction polymer chains. In the paper, studies on the stability of polymer solutions are described, an analysis of viscosity loss with time in the presence of iron and oxygen for polymer solutions is carried out. The choice of chemical deoxygenation method for the control of dissolved oxygen is substantiated. The program of laboratory studies of the ASP project and the analytical instruments used are described. The technological scheme of the ASP process is presented, and recommendations for the implementation of the technology are given.

  11. PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

    Science.gov (United States)

    Shor, R.S.; Vogler, S.

    1958-01-21

    A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

  12. Absorption of dissolved hydrogen from lithiated water during accelerated corrosion of zirconium-2.5 wt% niobium alloy

    International Nuclear Information System (INIS)

    Manolescu, A.V.; Mayer, P.; Rasile, E.M.; Mummenhoff, J.W.

    1982-01-01

    A series of laboratory experiments was carried out to determine the extent of dissolved hydrogen absorption from lithiated water by zirconium-2.5 wt% niobium alloy during corrosion. The material was exposed at 340 0 C to 1 M LiOH aqueous solution containing 0 to approximately 70 cm 3 /L of dissolved hydrogen. Results indicate that dissolved hydrogen has no effect on the corrosion rate or on the amount of hydrogen absorbed by the material

  13. Vibration analysis method for detection of abnormal movement of material in a rotary dissolver

    International Nuclear Information System (INIS)

    Smith, C.M.; Fry, D.N.

    1978-11-01

    Vibration signals generated by the movement of simulated nuclear fuel material through a three-stage, continuous, rotary dissolver were frequency analyzed to determine whether these signals contained characteristic signal patterns that would identify each of five phases of operation in the dissolver and, thus, would indicate the proper movement of material through the dissolver. This characterization of the signals is the first step in the development of a system for monitoring the flow of material through a dissolver to be developed for reprocessing spent nuclear fuel. Vibration signals from accelerometers mounted on the dissolver roller supports were analyzed in a bandwidth from 0 to 10 kHz. The analysis established that (1) all five phases of dissolver operation can be characterized by vibration signatures; (2) four of the five phases of operation can be readily and directly identified by a characteristic vibration signature during continuous, prototypic operation; (3) the transfer of material from the inlet to the dissolution stage can be indirectly monitored by one of the other four vibration signatures (the mixing signature) during prototypic operation; (4) a simulated blockage between the dissolution and exit stages can be detected by changes in one or more characteristic vibration signatures; and (5) a simulated blockage of the exit chute cannot be detected

  14. Thermotransport in interstitial solid solutions

    International Nuclear Information System (INIS)

    Fogel'son, R.L.

    1982-01-01

    On the basis of literature data the problem of thermotransport of impurities (H, N, O, C) in interstitial solid solutions is considered. It is shown that from experimental data on the thermotransport an important parameter of dissolved atoms can be found which characterizes atom state in these solutions-enthalpy of transport

  15. Dissolution of thorium/uranium mixed oxide in nitric acid-hydrofluoric acid solution

    International Nuclear Information System (INIS)

    Filgueiras, S.A.C.

    1984-01-01

    The dissolution process of thorium oxide and mixed uranium-thorium oxide is studied, as a step of the head-end of the fuel reprocessing. An extensive bibliography was analysed, concerning the main aspects of the system, specially the most important process variables. Proposed mechanisms and models for the thorium oxide dissolution are presented. The laboratory tests were performed in two phases: at first, powdered thoria was used as the material to be dissolved. The objective was to know how changes in he concentrations of the dissolvent solution components HNO 3 , HF and Al(NO 3 ) 3 affect the dissolution rate. The tests were planned according to the fractional factorial method. Thes results showed that it is advantageous to work with powdered material, since the reaction occurs rapidly. And, if the Thorex solution (HNO 3 13M, HF 0.05M and Al(NO 3 ) 3 0.10M) is a suitable dissolvent, it was verified that it is possible to reduce the concentration of either nitric or fluoridric acid, without reducing the reaction rate to an undesirable value. It was also observed significant interaction between the components of the dissolvent solution. In the second phase of the tests, (Th, 5%U)O 2 sintered pellets were used. The main goals were to know the pellets dissolution behaviour and to compare the results for different pellets among themselves. It was observed that the metallurgical history of the material strongly influences its dissolution, specially the density and the microstructure. It was also studied how the (Th,U)O 2 mass/Thorex solution volume ratio affects the time needed to obtain an 1 M Th/liter solution. The activation energy for the reaction was obtained. (Author) [pt

  16. Dynamics of dissolved and extractable organic nitrogen upon soil amendment with crop residues

    NARCIS (Netherlands)

    Ros, G.H.; Hoffland, E.

    2010-01-01

    Dissolved organic nitrogen (DON) is increasingly recognized as a pivotal pool in the soil nitrogen (N) cycle. Numerous devices and sampling procedures have been used to estimate its size, varying from in situ collection of soil solution to extraction of dried soil with salt solutions. Extractable

  17. PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

    Science.gov (United States)

    Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

    1962-08-14

    A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

  18. ANALYSIS OF DISSOLVED METHANE, ETHANE, AND ETHYLENE IN GROUND WATER BY A STANDARD GAS CHROMATOGRAPHIC TECHNIQUE

    Science.gov (United States)

    The measurement of dissolved gases such as methane, ethane, and ethylene in ground water is important in determining whether intrinsic bioremediation is occurring in a fuel- or solvent-contaminated aquifer. A simple procedure is described for the collection and subsequent analys...

  19. Influence of dissolved organic substances in groundwater on sorption behavior of americium and neptunium

    International Nuclear Information System (INIS)

    Boggs, S. Jr.; Seitz, M.G.

    1984-01-01

    Groundwaters typically contain dissolved organic carbon consisting largely of high molecular weight compounds of humic and fulvic acids. To evaluate whether these dissolved organic substances can enhance the tranport of radionuclides through the groundwater system, experiments were conducted to examine the sorption of americium and neptunium onto crushed basalt in the presence of dissolved humic- and fulvic-acid organic carbon introduced into synthetic groundwater. The partitioning experiments with synthetic groundwater show that increasing the concentration of either humic or fulvic acid in the water has a significant inhibiting effect on sorption of both americium and neptunium. At 22 0 C, adsorption of these radionuclides, as measured by distribution ratios (the ratio of nuclide sorbed onto the solid to nuclide in solution at the end of the experiment), decreased by 25% to 50% by addition of as little as 1 mg/L dissolved organic carbon and by one to two orders of magnitude by addition of 100 to 200 mg/L dissolved organic carbon. Distribution ratios measured in solutions reacted at 90 0 C similarly decreased with the addition of dissolved organic carbon but generally ranged from one to two orders of magnitude higher than those determined in the 22 0 C experiment. These results suggest that organic carbon dissolved in deep groundwaters may significantly enhance the mobility of radionuclides of americium and neptunium. 23 references, 5 figures, 11 tables

  20. Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

    International Nuclear Information System (INIS)

    Sivaraman, N.; Subramaniam, S.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2002-01-01

    Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel. (author)

  1. Subcooled boiling heat transfer correlation to calculate the effects of dissolved gas in a liquid

    International Nuclear Information System (INIS)

    Zarkasi, Amin S.; Chao, W.W.; Kunze, Jay F.

    2004-01-01

    The water coolant in most operating power reactor systems is kept free of dissolved gas, so as to minimize corrosion. However, in most research reactors, which operate at temperatures below 70 deg. C, and between 1 and 5 atm. pressure, the dissolved gas remains present in the water coolant system during operation. This dissolved gas can have a significant effect during accident conditions (i.e. a LOCA), when the fluid quickly reaches boiling, coincident with flow stagnation and subsequent flow reversal. A benchmark experiment was conducted, with an electrically heated, closed loop channel, modeling a research reactor fuel coolant channels (2 mm thick). The results showed 'boiling (bubble) noise' occurring before wall temperatures reached saturation, and a significant increase (up to 50%) in the heat transfer coefficient in the subcooled boiling region when in the presence of dissolved gas, compared to degassed water. Since power reactors do not involve dissolved gas, the RELAP safety analysis code does not include any provisions for the effect of dissolved gas on heat transfer. In this work, the effects of the dissolved gas are evaluated for inclusion in the RELAP code, including provision for initiating 'nucleate boiling' at a lower temperature, and a provision for enhancing the heat transfer coefficient during the subcooled boiling region. Instead of relying on Chen's correlation alone, a modification of the superposition method of Bjorge was adopted. (author)

  2. The iodine species and their behavior in the dissolution of spent-fuel specimens

    International Nuclear Information System (INIS)

    Sakurai, T.; Takahashi, A.; Ishikawa, N. Adachi, T.; Komaki, Y.; Ohnuki, M.

    1992-01-01

    In this paper, spent-fuel specimens (∼3 g each) with a burnup of 21 to 39 GWd/t were dissolved in 30 ml of 4 M HNO 3 at 100 degrees C, and the distribution of iodine and its chemical forms in the solution were studied. A small quantity of the iodine was conveyed to the insoluble residue (up to 2.3%), some remained in the fuel solution (up to 9.7%), and the balance was in the off-gas. Iodine was not deposited on the fuel cladding. Organic iodides were only ∼6.5% or less of the total amount of iodine in the off-gas. The fuel solution included iodine species that were difficult to expel by NO 2 sparging alone (27 to 46% of the iodine in the solution). These species were ascribed to be the colloids of AgI and PdI 2 . Iodate (IO - 3 ) was a rather minor iodine species in dissolution in ∼4 M HNO 3 . A thermochemical calculation also supports these results, indicating that the quantity of IO - 3 is ≤1.7 x 10 -4 % of the iodine fed to 4 M HNO 3 and that the colloid of AgI can be formed when the concentration of I - is ≥5.3 x 10 -10 M

  3. Criticality safety evaluation report for MKIA fuel pertaining to consolidating fuel storage

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1994-10-01

    Irradiated fuel criticality calculations are performed for MKIA fuel, 1.25 wt% 235 U, and 1.15 wt% 235 U fuel pieces and solutions. Comparisons are made between WIMS and MCNP. WIMS and MCNP calculations are documented

  4. Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

    International Nuclear Information System (INIS)

    Gray, W.J.; Einziger, R.E.

    1998-04-01

    Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

  5. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  6. Application of the gravimetric method to closing the material balance around the chop-leach cell of a spent-fuel reprocessing plant

    International Nuclear Information System (INIS)

    Fishbone, L.G.

    1985-01-01

    For a spent-fuel reprocessing plant handling commercial light-water-reactor fuel, plutonium accounting is traditionally done for the material balance area (MBA) extending from the input accountability tank to the product accountability tank - the process MBA. Consider an MBA comprising the chop-leach cell, with an inward flow consisting of the intact spent-fuel assemblies and outward flows consisting of leached hulls and dissolver solution. Given knowledge of the original uranium mass in the fuel and a measurement of the uranium-plutonium concentration ratio in the dissolver solution, the gravimetric method can be used to determine the amount of plutonium in the spent-fuel assemblies. A measurement of residual plutonium in the leached hulls would then permit the determination of a plutonium material balance for the chop-leach cell alone, since the volumetrically determined plutonium in the input accountability tank yields the plutonium in the flow leaving the chop-leach cell for the process MBA. The uncertainty in the balance can be estimated given the individual measurement uncertainties

  7. Water reactive hydrogen fuel cell power system

    Science.gov (United States)

    Wallace, Andrew P; Melack, John M; Lefenfeld, Michael

    2014-01-21

    A water reactive hydrogen fueled power system includes devices and methods to combine reactant fuel materials and aqueous solutions to generate hydrogen. The generated hydrogen is converted in a fuel cell to provide electricity. The water reactive hydrogen fueled power system includes a fuel cell, a water feed tray, and a fuel cartridge to generate power for portable power electronics. The removable fuel cartridge is encompassed by the water feed tray and fuel cell. The water feed tray is refillable with water by a user. The water is then transferred from the water feed tray into a fuel cartridge to generate hydrogen for the fuel cell which then produces power for the user.

  8. Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.

    1985-01-01

    Onset of melting is an important performance limit for irradiated UO 2 and UO 2 -based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO 2 and (U,PU)O 2 , however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO 2 -PUO 2 , UO 2 -CeO 2 , UO 2 -BaO, UO 2 -SrO, UO 2 -BaZrO 3 and UO 2 -SrZrO 3 . These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies. (orig.)

  9. Analysis of an homogeneous solution reactor for 99 Mo production

    International Nuclear Information System (INIS)

    Weir, A.; Lopasso, E.; Gho, C.

    2007-01-01

    The 99m Tc is the more used radioisotope in nuclear medicine, used in 80% of procedures of nuclear medicine in the world. This is due to their characteristics practically ideal for the diagnostic. The 99m Tc is obtained by decay of the 99 Mo, which can produce it by irradiating enriched targets in 98 Mo, or as fission product, irradiating uranium targets or by means of homogeneous solution reactors. The pattern of the used reactor in the neutron analysis possesses a liquid fuel composed of uranyl nitrate dissolved in water with the attach of nitric acid. This solution is contained in a cylindrical recipient of stainless steel reflected with light water. The reactor is refrigerated by means of an helicoidal heat exchanger immersed in the fuel solution. The heat of the fuel is removed by natural convection while the circulation of the water inside the exchanger is forced. The control system of the reactor consists on 6 independent cadmium bars, with followers of water. An auxiliary control system can be the level of the fuel solution inside container tank, but it was not included in the pattern in study. One studies the variations of the reactivity of the system due to different phenomena. An important factor during the normal operation of the reactor is the variation of temperature taking to a volumetric expansion of the fuel and ghastly effects in the same one. Another causing phenomenon of changes in the reactivity is the variation of the concentration of uranium in the combustible solution. An important phenomenon in this type of reactors is the hole fraction in the nucleus I liquidate due to the radiolysis and the possible boil of the water of the combustible solution. Some of the possible cases of abnormal operation were studied as the lost one of coolant in the secondary circuit of the heat exchanger, the introduction and evaporation of water in the nucleus. The reactivity variations were studied using the codes of I calculate MCNP, WIMS and TORT. All the

  10. Simultaneous effect of dissolved organic carbon, surfactant, and organic acid on the desorption of pesticides investigated by response surface methodology

    DEFF Research Database (Denmark)

    Trinh, Ha Thu; Duong, Hanh Thi; Ta, Thao Thi

    2017-01-01

    Desorption of pesticides (fenobucarb, endosulfan, and dichlorodiphenyltrichloroethane (DDT)) from soil to aqueous solution with the simultaneous presence of dissolved organic carbon (DOC), sodium dodecyl sulfate (SDS), and sodium oxalate (Oxa) was investigated in batch test by applying a full...

  11. Nuclear fuel technology - Determination of milligram amounts of plutonium in nitric acid solutions - Potentiometric titration with potassium dichromate after oxidation by Ce(IV) and reduction by Fe(II)

    International Nuclear Information System (INIS)

    2000-01-01

    This International Standard describes a precise and accurate analytical method for determining 1 mg to 5 mg of plutonium per millilitre in nitric acid solutions. The method is very selective for plutonium. It is suitable for the direct determination of plutonium in materials ranging from pure product solutions, to solutions of mixed nuclear materials with a uranium/plutonium ratio up to 20:1. However, potential application to the assay of plutonium in solutions of irradiated nuclear fuels and solutions of mixed nuclear materials with uranium/plutonium ratios of 20:1 to 33:1 has not yet been documented. The method recommends that the aliquot be weighed and that the titration burettes be calibrated gravimetrically in order to obtain adequate precision and accuracy. This does not preclude using any alternative technique which can be shown to give an equivalent accuracy. As the reproducibility of the reaction conditions is important to maintain good performance, extensive automatization of the procedure is beneficial

  12. Simulation of facility operations and materials accounting for a combined reprocessing/MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Coulter, C.A.; Whiteson, R.; Zardecki, A.

    1991-01-01

    We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs

  13. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  14. Cycling downwards - dissolved organic matter in soils

    NARCIS (Netherlands)

    Kaiser, K.; Kalbitz, K.

    2012-01-01

    Dissolved organic matter has been recognized as mobile, thus crucial to translocation of metals, pollutants but also of nutrients in soil. We present a conceptual model of the vertical movement of dissolved organic matter with soil water, which deviates from the view of a chromatographic stripping

  15. Regeneration of porous nickel elements. [an aqueous solution of NH/sub 3/--NH/sub 4/Cl is passed through cell to remove nickel oxides

    Energy Technology Data Exchange (ETDEWEB)

    Winsel, A; Von Doehren, H H

    1972-01-27

    A method for regenerating a fuel cell with Ag-catalyzed O electrodes containing Ni and H electrodes containing Raney Ni where the voltage had dropped from 750 to 630 mV within 3200 hr at 50 mA/cm/sup 2/ is described. An aqueous NH/sub 3/-NH/sub 4/Cl solution was passed through the cell under 1 atm H at 60/sup 0/, whereby 27 g Ni was dissolved as the hydroxide. The voltage of the regenerated cell was 770 mV and remained constant during 500 hr operation. The Ni ions in the regenerating solutions were removed by electrolysis.

  16. Spectrophotometric determination of dissolved tri n-butyl phosphate in aqueous streams of Purex process

    International Nuclear Information System (INIS)

    Ganesh, S.; Velavendan, P.; Pandey, N.K.; Ahmed, M.K.; Kamachi Mudali, U.; Natarajan, R.

    2012-01-01

    A spectrophotometric method is developed for the determination of dissolved tri-n butyl phosphate (TBP) in aqueous streams of Purex process used in nuclear fuel reprocessing. The method is based on the formation of phosphomolybdate with added ammonium molybdate followed by reduction with hydrazine sulphate in acid medium. Orthophosphate and molybdate ions combine in acidic solution to give molybdophosphoric (phosphomolybdic) acid, which upon selective reduction (with hydrazinium sulphate) produces a blue colour, due to molybdenum blue. The intensity of blue colour is proportional to the amount of phosphate. If the acidity at the time of reduction is 0.5 M in sulphuric acid and hydrazinium sulphate is the reductant, the resulting blue complex exhibits maximum absorption at 810-840 nm. The system obeys Lambert-Beer's law at 830 nm in the concentration range of 0.1-1.0 μg/mol of phosphate. Molar Absorptivity was determined to be 3.1 x 10 4 L mol -1 cm -1 at 830 nm. The results obtained are reproducible with standard deviation of 1 % and relative error less than 2 % and are in good agreement with those obtained by ion chromatographic technique. (author)

  17. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-01-01

    The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

  18. Dissolved air flotation and me.

    Science.gov (United States)

    Edzwald, James K

    2010-04-01

    This paper is mainly a critical review of the literature and an assessment of what we know about dissolved air flotation (DAF). A few remarks are made at the outset about the author's personal journey in DAF research, his start and its progression. DAF has been used for several decades in drinking water treatment as an alternative clarification method to sedimentation. DAF is particularly effective in treating reservoir water supplies; those supplies containing algae, natural color or natural organic matter; and those with low mineral turbidity. It is more efficient than sedimentation in removing turbidity and particles for these type supplies. Furthermore, it is more efficient in removing Giardia cysts and Cryptosporidium oocysts. In the last 20 years, fundamental models were developed that provide a basis for understanding the process, optimizing it, and integrating it into water treatment plants. The theories were tested through laboratory and pilot-plant studies. Consequently, there have been trends in which DAF pretreatment has been optimized resulting in better coagulation and a decrease in the size of flocculation tanks. In addition, the hydraulic loading rates have increased reducing the size of DAF processes. While DAF has been used mainly in conventional type water plants, there is now interest in the technology as a pretreatment step in ultrafiltration membrane plants and in desalination reverse osmosis plants. Copyright (c) 2009 Elsevier Ltd. All rights reserved.

  19. Particle Swarm Optimization applied to combinatorial problem aiming the fuel recharge problem solution in a nuclear reactor; Particle swarm optimization aplicado ao problema combinatorio com vistas a solucao do problema de recarga em um reator nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: ameneses@con.ufrj.br; schirru@lmp.ufrj.br

    2005-07-01

    This work focuses on the usage the Artificial Intelligence technique Particle Swarm Optimization (PSO) to optimize the fuel recharge at a nuclear reactor. This is a combinatorial problem, in which the search of the best feasible solution is done by minimizing a specific objective function. However, in this first moment it is possible to compare the fuel recharge problem with the Traveling Salesman Problem (TSP), since both of them are combinatorial, with one advantage: the evaluation of the TSP objective function is much more simple. Thus, the proposed methods have been applied to two TSPs: Oliver 30 and Rykel 48. In 1995, KENNEDY and EBERHART presented the PSO technique to optimize non-linear continued functions. Recently some PSO models for discrete search spaces have been developed for combinatorial optimization. Although all of them having different formulation from the ones presented here. In this paper, we use the PSO theory associated with to the Random Keys (RK)model, used in some optimizations with Genetic Algorithms. The Particle Swarm Optimization with Random Keys (PSORK) results from this association, which combines PSO and RK. The adaptations and changings in the PSO aim to allow the usage of the PSO at the nuclear fuel recharge. This work shows the PSORK being applied to the proposed combinatorial problem and the obtained results. (author)

  20. A highly accurate method for determination of dissolved oxygen: Gravimetric Winkler method

    International Nuclear Information System (INIS)

    Helm, Irja; Jalukse, Lauri; Leito, Ivo

    2012-01-01

    Highlights: ► Probably the most accurate method available for dissolved oxygen concentration measurement was developed. ► Careful analysis of uncertainty sources was carried out and the method was optimized for minimizing all uncertainty sources as far as practical. ► This development enables more accurate calibration of dissolved oxygen sensors for routine analysis than has been possible before. - Abstract: A high-accuracy Winkler titration method has been developed for determination of dissolved oxygen concentration. Careful analysis of uncertainty sources relevant to the Winkler method was carried out and the method was optimized for minimizing all uncertainty sources as far as practical. The most important improvements were: gravimetric measurement of all solutions, pre-titration to minimize the effect of iodine volatilization, accurate amperometric end point detection and careful accounting for dissolved oxygen in the reagents. As a result, the developed method is possibly the most accurate method of determination of dissolved oxygen available. Depending on measurement conditions and on the dissolved oxygen concentration the combined standard uncertainties of the method are in the range of 0.012–0.018 mg dm −3 corresponding to the k = 2 expanded uncertainty in the range of 0.023–0.035 mg dm −3 (0.27–0.38%, relative). This development enables more accurate calibration of electrochemical and optical dissolved oxygen sensors for routine analysis than has been possible before.

  1. Standard method of test for atom percent fission in uranium fuel - radiochemical method

    International Nuclear Information System (INIS)

    Anon.

    The determination of the U at. % fission that has occurred in U fuel from an analysis of the 137 Cs ratio to U ratio after irradiation is described. The method is applicable to high-density, clad U fuels (metal, alloys, or ceramic compounds) in which no separation of U and Cs has occurred. The fuels are best aged for several months after irradiation in order to reduce the 13-day 136 Cs activity. The fuel is dissolved and diluted to produce a solution containing a final concentration of U of 100 to 1000 mg U/l. The 137 Cs concentration is determined by ASTM method E 320, for Radiochemical Determination of Cesium-137 in Nuclear Fuel Solutions, and the U concentration is determined by ASTM method E 267, for Determination of Uranium and Plutonium Concentrations and Isotopic Abundances, ASTM method E 318, for Colorimetric Determination of Uranium by Controlled-Potential Coulometry. Calculations are given for correcting the 137 Cs concentration for decay during and after irradiation. The accuracy of this method is limited, not only by the experimental errors with which the fission yield and the half-life of 137 Cs are known

  2. Kinetics of two phase fuel reflected reactors

    International Nuclear Information System (INIS)

    Buzano, M.L.; Corno, S.E.; Mattioda, F.

    2000-01-01

    In the present work a self-consistent mathematical model for the local dynamics of a quite particular class of fission reactors has been developed and solved. These devices consist of an innermost multiplying region, in which a significant fraction of the fissile fuel is diluted into a liquid phase, while the complementary fuel fraction operates as a standing solid matrix. This unconventional active region is surrounded by a standard peripheral reflector. For cooling purposes, the fluid fraction of the fuel needs to be circulated through external heat exchangers. The pump driven circulation causes the delayed neutron precursors, dissolved inside the fluid phase, to be spatially homogenized in the core volume well before decaying, while a continuous removal of precursor nuclei from the core takes place as a consequence of the outside circulation. Furthermore, the fraction of the extracted precursors still surviving after the solenoidal trip through the heat exchangers is continuously reinserted into the core. A new type of dynamical model is required to account for these unusual technological features. The mathematical structure of the evolution model presented in this paper consists of a system of integro-differential-difference equations, whose solution is derived in closed-form, by means of fully analytical techniques. Many dynamics and safety features of reactors of this type can be clarified a priori, upon inspection of the mathematical properties of the solution of the model. The rigorous time-eigenvalue generating equation can be explicitly established in the present theoretical context, together with the evaluation of any kind of transients. A short survey on the possible fields of application of these reactors is also presented

  3. Iron traps terrestrially derived dissolved organic matter at redox interfaces

    Science.gov (United States)

    Riedel, Thomas; Zak, Dominik; Biester, Harald; Dittmar, Thorsten

    2013-01-01

    Reactive iron and organic carbon are intimately associated in soils and sediments. However, to date, the organic compounds involved are uncharacterized on the molecular level. At redox interfaces in peatlands, where the biogeochemical cycles of iron and dissolved organic matter (DOM) are coupled, this issue can readily be studied. We found that precipitation of iron hydroxides at the oxic surface layer of two rewetted fens removed a large fraction of DOM via coagulation. On aeration of anoxic fen pore waters, >90% of dissolved iron and 27 ± 7% (mean ± SD) of dissolved organic carbon were rapidly (within 24 h) removed. Using ultra-high-resolution MS, we show that vascular plant-derived aromatic and pyrogenic compounds were preferentially retained, whereas the majority of carboxyl-rich aliphatic acids remained in solution. We propose that redox interfaces, which are ubiquitous in marine and terrestrial settings, are selective yet intermediate barriers that limit the flux of land-derived DOM to oceanic waters. PMID:23733946

  4. Dissolved CO2 Increases Breakthrough Porosity in Natural Porous Materials.

    Science.gov (United States)

    Yang, Y; Bruns, S; Stipp, S L S; Sørensen, H O

    2017-07-18

    When reactive fluids flow through a dissolving porous medium, conductive channels form, leading to fluid breakthrough. This phenomenon is caused by the reactive infiltration instability and is important in geologic carbon storage where the dissolution of CO 2 in flowing water increases fluid acidity. Using numerical simulations with high resolution digital models of North Sea chalk, we show that the breakthrough porosity is an important indicator of dissolution pattern. Dissolution patterns reflect the balance between the demand and supply of cumulative surface. The demand is determined by the reactive fluid composition while the supply relies on the flow field and the rock's microstructure. We tested three model scenarios and found that aqueous CO 2 dissolves porous media homogeneously, leading to large breakthrough porosity. In contrast, solutions without CO 2 develop elongated convective channels known as wormholes, with low breakthrough porosity. These different patterns are explained by the different apparent solubility of calcite in free drift systems. Our results indicate that CO 2 increases the reactive subvolume of porous media and reduces the amount of solid residual before reactive fluid can be fully channelized. Consequently, dissolved CO 2 may enhance contaminant mobilization near injection wellbores, undermine the mechanical sustainability of formation rocks and increase the likelihood of buoyance driven leakage through carbonate rich caprocks.

  5. Analysis of a slow-dissolving medicine by EPMA

    International Nuclear Information System (INIS)

    Sasayama, Tetsuaki; Kohara, Kiyohiro; Araki, Takeshi

    1995-01-01

    Along with a dissolution test of a slow-dissolving medicine, the change in distribution of the drug in solution can be observed by using EPMA, and the structual factors and dissolution mechanism which determine the bioavailability of medicine can be clarified. In the evaluation of physical, chemical and pharmaceutical qualities, it is concluded that EPMA is very effective in elemental and state analyses with observation of microscopic areas on the micrometer order. Especially, the color mapping method clarifies the distribution of a drug in the total image field and enables us to analyze the mechanism of a dissolution medicine. (author)

  6. Measurement of total dissolved solids using electrical conductivity

    International Nuclear Information System (INIS)

    Ray, Vinod K.; Jat, J.R.; Reddy, G.B.; Balaji Rao, Y.; Phani Babu, C.; Kalyanakrishnan, G.

    2017-01-01

    Total dissolved solids (TDS) is an important parameter for the disposal of effluents generated during processing of different raw materials like Magnesium Di-uranate (MDU), Heat Treated Uranium Peroxide (HTUP), Sodium Di-uranate (SDU) in Uranium Extraction plant and Washed and Dried Frit (WDF) in Zirconium Extraction Plant. The present paper describes the use of electrical conductivity for determination of TDS. As electrical conductivity is matrix dependent property, matrix matched standards were prepared for determination of TDS in ammonium nitrate solution (AN) and mixture of ammonium nitrate and ammonium sulphate (AN/AS) and results were found to be in good agreement when compared with evaporation method. (author)

  7. Measurement and interpretation of low levels of dissolved oxygen in ground water

    Science.gov (United States)

    White, A.F.; Peterson, M.L.; Solbau, R.D.

    1990-01-01

    A Rhodazine-D colorimetric technique was adapted to measure low-level dissolved oxygen concentrations in ground water. Prepared samples containing between 0 and 8.0 ??moles L-1 dissolved oxygen in equilibrium with known gas mixtures produced linear spectrophotometric absorbance with a lower detection limit of 0.2 ??moles L-1. Excellent reproducibility was found for solutions ranging in composition from deionized water to sea water with chemical interferences detected only for easily reduced metal species such as ferric ion, cupric ion, and hexavalent chromium. Such effects were correctable based on parallel reaction stoichiometries relative to oxygen. The technique, coupled with a downhole wire line tool, permitted low-level monitoring of dissolved oxygen in wells at the selenium-contaminated Kesterson Reservoir in California. Results indicated a close association between low but measurable dissolved oxygen concentrations and mobility of oxidized forms of selenium. -from Authors

  8. Millstone 3 condensate dissolved gas monitoring

    International Nuclear Information System (INIS)

    Burns, T.F.; Grondahl, E.E.; Snyder, D.T.

    1988-01-01

    Condensate dissolved oxygen problems at Millstone Point Unit 3 (MP3) were investigated using the Dissolved Gas Monitoring System developed by Radiological and Chemical Technology, Inc. under EPRI sponsorship. Argon was injected into the turbine exhaust basket tips to perform a dissolved gas transport analysis and determine steam jet air ejector gas removal efficiency. The operating configuration of the steam jet air ejector system was varied to determine the effect on gas removal efficiency. Following circulating water chlorination, the gas removal efficiency was determined to evaluate the effect of condenser tube fouling on steam jet air ejector performance

  9. The effect of dissolved oxygen on water radiolysis behaviour

    International Nuclear Information System (INIS)

    Yakabuskie, P.A.; Joseph, J.M.; Wren, J.C.; Stuart, C.R.

    2012-09-01

    A quantitative understanding of the chemical or redox environments generated in water by ionizing radiation is important for material selection, development of maintenance programs, and safety assessments for water-cooled nuclear power reactors. The highly reactive radicals (·OH, ·H, ·e aq - , ·HO 2 , and ·O 2 - ) and molecular species (H 2 and H 2 O 2 ) generated by water radiolysis can compete in reactions with other dissolved compounds and impose changes to the system chemistry by altering the steady-state concentrations of water radiolysis products, which could impact the degradation of materials in contact with the aqueous phase. Understanding in detail how a given chemical additive changes the long-term radiolysis kinetics can help us to determine what chemistry control steps may be required to return the system to an optimal redox condition, and in turn, enhance the lifetime of reactor components. This study outlines the effect of dissolved oxygen gas, which could be introduced due to air ingress, on long-term water radiolysis behaviour. The effects of solution pH and initial dissolved O 2 concentration on the radiolytic production of molecular H 2 and H 2 O 2 have been investigated by performing experiments with three different O 2 concentrations at pH 6.0 and 10.6 under steady-state radiolysis conditions. The aqueous and gas phase analyses were performed using UV-Vis spectrophotometry and gas-chromatography equipped with electron capture and thermal conductivity detectors. The experimental results were compared with kinetic model calculations of steady-state radiolysis and were found to be in good agreement. The concentrations of water radiolysis products, H 2 O 2 and H 2 , were found to increase in the presence of dissolved oxygen, but the degree of increase was shown to depend on the solution pH. Furthermore, the steady-state concentration of H 2 did not increase as greatly as that of H 2 O 2 at either pH studied. The kinetic analyses have shown

  10. Nuclear fuel management via fuel quality factor averaging

    International Nuclear Information System (INIS)

    Mingle, J.O.

    1978-01-01

    The numerical procedure of prime number averaging is applied to the fuel quality factor distribution of once and twice-burned fuel in order to evolve a fuel management scheme. The resulting fuel shuffling arrangement produces a near optimal flat power profile both under beginning-of-life and end-of-life conditions. The procedure is easily applied requiring only the solution of linear algebraic equations. (author)

  11. Convective mass transfer around a dissolving bubble

    Science.gov (United States)

    Duplat, Jerome; Grandemange, Mathieu; Poulain, Cedric

    2017-11-01

    Heat or mass transfer around an evaporating drop or condensing vapor bubble is a complex issue due to the interplay between the substrate properties, diffusion- and convection-driven mass transfer, and Marangoni effects, to mention but a few. In order to disentangle these mechanisms, we focus here mainly on the convective mass transfer contribution in an isothermal mass transfer problem. For this, we study the case of a millimetric carbon dioxide bubble which is suspended under a substrate and dissolved into pure liquid water. The high solubility of CO2 in water makes the liquid denser and promotes a buoyant-driven flow at a high (solutal) Rayleigh number (Ra˜104 ). The alteration of p H allows the concentration field in the liquid to be imaged by laser fluorescence enabling us to measure both the global mass flux (bubble volume, contact angle) and local mass flux around the bubble along time. After a short period of mass diffusion, where the boundary layer thickens like the square root of time, convection starts and the CO2 is carried by a plume falling at constant velocity. The boundary layer thickness then reaches a plateau which depends on the bubble cross section. Meanwhile the plume velocity scales like (dV /d t )1 /2 with V being the volume of the bubble. As for the rate of volume loss, we recover a constant mass flux in the diffusion-driven regime followed by a decrease in the volume V like V2 /3 after convection has started. We present a model which agrees well with the bubble dynamics and discuss our results in the context of droplet evaporation, as well as high Rayleigh convection.

  12. Measurement of the hydrogen yield in the radiolysis of water by dissolved fission products

    International Nuclear Information System (INIS)

    Sauer, M.C. Jr.; Hart, E.J.; Flynn, K.F.; Gindler, J.E.

    1976-04-01

    Hydrogen from the radiolysis of water by dissolved fission products is stripped from the solution and collected by bubbling CO 2 through the solution. Quantitative measurements of the G value for hydrogen show that the yield is essentially the same as would be obtained by external gamma radiolysis of nonradioactive solutions of the same chemical composition. The hydrogen yield can be enhanced by addition of a hydrogen-atom donor, such as formic acid, to the solution. The yield of hydrogen from fission-waste solutions is discussed with respect to the question of whether it represents a significant energy source

  13. Behavior of iodine in the dissolution of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Tsutomu; Komatsu, Kazunori; Takahashi, A. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    The results of laboratory-scale experiments concerning the behavior of iodine in the dissolution of spent nuclear fuels, which were carried out at the Japan Atomic Energy Research Institute, are summarized. Based on previous and new experimental results, the difference in quantity of residual iodine in the fuel solution between laboratory-scale experiments and reprocessing plants is discussed, Iodine in spent fuels is converted to the following four states: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid generated in the dissolution, (3) formation of a colloid of insoluble iodides such as AgI and PdI{sub 2}, and (4) deposition on insoluble residue. Nitrous acid controls the amount of colloid formed. As a result, up to 10% of iodine in spent fuels is retained in the fuel solution, up to 3% is deposited on insoluble residue, and the balance volatilizes to the off-gas, Contrary to earlier belief, when the dissolution is carried out in 3 to 4 M HNO{sub 3} at 100{degrees}C, the main iodine species in a fuel solution is a colloid, not iodate, Immediately after its formation, the colloid is unstable and decomposes partially in the hot nitric acid solution through the following reaction: AgI(s) + 2HNO{sub 3}(aq) = {1/2}I{sub 2}(aq) + AgNO{sub 3}(aq) + NO{sub 2}(g) + H{sub 2}O(1). For high concentrations of gaseous iodine, I{sub 2}(g), and NO{sub 2}, this reaction is reversed towards formation of the colloid (AgI). Since these concentrations are high near the liquid surface of a plant-scale dissolver, there is a possibility that the colloid is formed there through this reversal, Simulations performed in laboratory-scale experiments demonstrated this reversal, This phenomenon can be one reason the quantity of residual iodine in spent fuels is higher in reprocessing plants than in laboratory-scale experiments. 17 refs., 5 figs., 3 tabs.

  14. Comparative Emulsifying Properties of Octenyl Succinic Anhydride (OSA-Modified Starch: Granular Form vs Dissolved State.

    Directory of Open Access Journals (Sweden)

    María Matos

    Full Text Available The emulsifying ability of OSA-modified and native starch in the granular form, in the dissolved state and a combination of both was compared. This study aims to understand mixed systems of particles and dissolved starch with respect to what species dominates at droplet interfaces and how stability is affected by addition of one of the species to already formed emulsions. It was possible to create emulsions with OSA-modified starch isolated from Quinoa as sole emulsifier. Similar droplet sizes were obtained with emulsions prepared at 7% (w/w oil content using OSA-modified starch in the granular form or molecularly dissolved but large differences were observed regarding stability. Pickering emulsions kept their droplet size constant after one month while emulsions formulated with OSA-modified starch dissolved exhibited coalescence. All emulsions stabilized combining OSA-modified starch in granular form and in solution showed larger mean droplet sizes with no significant differences with respect to the order of addition. These emulsions were unstable due to coalescence regarding presence of free oil. Similar results were obtained when emulsions were prepared by combining OSA-modified granules with native starch in solution. The degree of surface coverage of starch granules was much lower in presence of starch in solution which indicates that OSA-starch is more surface active in the dissolved state than in granular form, although it led to unstable systems compared to starch granule stabilized Pickering emulsions, which demonstrated to be extremely stable.

  15. Advanced oxidation of biorefractory organics in aqueous solution together with bioelectricity generation by microbial fuel cells with composite FO/GPEs

    Science.gov (United States)

    Fu, Bao-rong; Shen, Chao; Ren, Jing; Chen, Jia-yi; Zhao, Lin

    2018-03-01

    In this study, ferric oxide loading graphite particle electrodes (FO/GPEs) were prepared as cathode of a three-dimensional electrode MFC-Fenton system. The properties of the composite cathode were examined with higher surface area and more mesopores. FO/GPEs could work as both cathode and Fenton iron reagents, contributing to high oxidation activity and better performance of electricity generation. The application of FO/GPEs MFC-Fenton system on degrading p-nitrophenol presented high catalytic efficiency in a wide range of pH value. The removal of p-nitrophenol and TOC attained to about 85 % within 8 and 64 h at neutral pH, respectively. A neutral FO/GPEs MFC-Fenton oxidation mechanism was also proposed. Specifically, both the surface iron sites and dissolved iron ions catalyzed the decomposition of H2O2. As results, the generated hydroxyl radicals were used for p-nitrophenol degradation and the iron oxide was recycled.

  16. Is Yucca Mountain a long-term solution for disposing of US spent nuclear fuel and high-level radioactive waste?

    Science.gov (United States)

    Thorne, M C

    2012-06-01

    On 26 January 2012, the Blue Ribbon Commission on America's Nuclear Future released a report addressing, amongst other matters, options for the managing and disposal of high-level waste and spent fuel. The Blue Ribbon Commission was not chartered as a siting commission. Accordingly, it did not evaluate Yucca Mountain or any other location as a potential site for the storage or disposal of spent nuclear fuel and high-level waste. Nevertheless, if the Commission's recommendations are followed, it is clear that any future proposals to develop a repository at Yucca Mountain would require an extended period of consultation with local communities, tribes and the State of Nevada. Furthermore, there would be a need to develop generally applicable regulations for disposal of spent fuel and high-level radioactive waste, so that the Yucca Mountain site could be properly compared with alternative sites that would be expected to be identified in the initial phase of the site-selection process. Based on what is now known of the conditions existing at Yucca Mountain and the large number of safety, environmental and legal issues that have been raised in relation to the DOE Licence Application, it is suggested that it would be imprudent to include Yucca Mountain in a list of candidate sites for future evaluation in a consent-based process for site selection. Even if there were a desire at the local, tribal and state levels to act as hosts for such a repository, there would be enormous difficulties in attempting to develop an adequate post-closure safety case for such a facility, and in showing why this unsaturated environment should be preferred over other geological contexts that exist in the USA and that are more akin to those being studied and developed in other countries.

  17. Dissolving microneedle patches for dermal vaccination

    OpenAIRE

    Leone, M.; Monkare, J.T.; Bouwstra, J.A.; Kersten, G.F.A.

    2017-01-01

    The dermal route is an attractive route for vaccine delivery due to the easy skin accessibility and a dense network of immune cells in the skin. The development of microneedles is crucial to take advantage of the skin immunization and simultaneously to overcome problems related to vaccination by conventional needles (e.g. pain, needle-stick injuries or needle re-use). This review focuses on dissolving microneedles that after penetration into the skin dissolve releasing the encapsulated antige...

  18. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  19. Optimization method development of the core characteristics of a fast reactor in order to explore possible high performance solutions (a solution being a consistent set of fuel, core, system and safety)

    International Nuclear Information System (INIS)

    Ingremeau, J.-J.X.

    2011-01-01

    In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc...) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) uses analytical models and interpolations (Meta-models) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a 'buffer', a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that

  20. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  1. Method for improving solution flow in solution mining of a mineral

    International Nuclear Information System (INIS)

    Moore, T.

    1980-01-01

    An improved method for the solution mining of a mineral from a subterranean formation containing same in which an injection and production well are drilled and completed within said formation, leach solution and an oxidant are injected through said injection well into said formation to dissolve said mineral, and said dissolved mineral is recovered via said production well, wherein the improvement comprises pretreating said formation with an acid gas to improve the permeabiltiy thereof

  2. The effect of the oxygen dissolved in the adsorption of gold in activated carbon

    International Nuclear Information System (INIS)

    Navarro, P.; Wilkomirsky, I.

    1999-01-01

    The effect of the oxygen dissolved on the adsorption of gold in a activated carbon such as these used for carbon in pulp (CIP) and carbon in leach (CIL) processes were studied. The research was oriented to dilucidate the effect of the oxygen dissolved in the gold solution on the kinetics and distribution of the gold adsorbed in the carbon under different conditions of ionic strength, pH and gold concentration. It was found that the level of the oxygen dissolved influences directly the amount of gold adsorbed on the activated carbon, being this effect more relevant for low ionic strength solutions. The pH and initial gold concentration has no effect on this behavior. (Author) 16 refs

  3. Capacitance evolution of electrochemical capacitors with tailored nanoporous electrodes in pure and dissolved ionic liquids

    Energy Technology Data Exchange (ETDEWEB)

    Mysyk, R.; Raymundo-Pinero, E. [CRMD, CNRS/University, Orleans (France); Ruiz, V.; Santamaria, R. [Instituto Nacional del Carbon (CSIC), Oviedo (Spain); Beguin, F.

    2010-10-15

    A homologous series of ionic liquids (IL) with 1-alkyl-3-methylimidazolium cations of different lengths of alkyl chain was used to study the effect of cation size on the capacitive response of two carbons with a tailored pore size distribution. The results reveal a clear ion-sieving effect in pure ILs, while the effect is heavily mitigated for the same salts used in solution, most likely due to somewhat stronger geometrical flexibility of dissolved ions. For the electrode material showing the ion-sieving effect in solution, the gravimetric capacitance values are higher than in pure ILs. The dissimilarity of capacitance values between pure and dissolved ILs with ion-sieving carbons highlights their respective advantages and disadvantages in terms of energy density: whereas pure ILs can potentially provide a larger working voltage window, the corresponding dissolved salts can access smaller pores, mostly contributing to higher capacitance values. (Abstract Copyright [2010], Wiley Periodicals, Inc.)

  4. Fabrication of ATALANTE Dissolver Off-Gas Sorbent-Based Capture System

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Jr., Joseph Franklin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    A small sorbent-based capture system was designed that could be placed in the off-gas line from the fuel dissolver in the ATALANTE hot cells with minimal modifications to the ATALANTE dissolver off-gas system. Discussions with personnel from the ATALANTE facility provided guidance that was used for the design. All components for this system have been specified, procured, and received on site at Oak Ridge National Laboratory (ORNL), meeting the April 30, 2015, milestone for completing the fabrication of the ATALANTE dissolver off-gas capture system. This system will be tested at ORNL to verify operation and to ensure that all design requirements for ATALANTE are met. Modifications to the system will be made, as indicated by the testing, before the system is shipped to ATALANTE for installation in the hot cell facility.

  5. Effects of dissolved species on radiolysis of diluted seawater

    International Nuclear Information System (INIS)

    Hata, Kuniki; Hanawa, Satoshi; Kasahara, Shigeki; Motooka, Takafumi; Tsukada, Takashi; Muroya, Yusa; Yamashita, Shinichi; Katsumura, Yosuke

    2014-01-01

    Fukushima Daiichi Nuclear Power Plants (NPPs) experienced seawater injection into the cores and fuel pools as an emergent measure after the accident. After the accident, retained water has been continuously desalinized, and subsequently the concentration of chloride ion (Cl"-) has been kept at a lower level these days. These ions in seawater are known to affect water radiolysis, which causes the production of radiolytic products, such as hydrogen peroxide (H_2O_2), molecular hydrogen (H_2) and molecular oxygen (O_2). However, the effects of dissolved ions relating seawater on the production of the stable radiolytic products are not well understood in the diluted seawater. To understand of the production behavior in diluted seawater under radiation, radiolysis calculations were carried out. Production of H_2 is effectively suppressed by diluting by up to vol10%. The concentrations of oxidants (H_2O_2 and O_2) are also suppressed by dilution of dissolved species. The effect of oxidants on corrosion of materials is thought to be low when the seawater was diluted by less than 1 vol% by water. It is also shown that deaeration is one of the effective measure to suppress the concentrations of oxidants at a lower level for any dilution conditions. (author)

  6. Separation of radio cesium from PUREX feed solution by sorption on composite ammonium molybdo phosphate (AMP)

    International Nuclear Information System (INIS)

    Singh, I.J.; Achuthan, P.V.; Jain, S.; Janardanan, C.; Gopalakrishnan, V.; Wattal, P.K.; Ramanujam, A.

    2001-01-01

    Composite AMP exchanger was developed and evaluated for separation of radio cesium from dissolver solutions of PUREX process using a column experiment. The composite shows excellent sorption of radio cesium from dissolver solutions without any loss of plutonium and uranium. The removal of radio cesium from dissolver solutions will help in lowering the degradation of tri-n-butyl phosphate (TBP) in the solvent extraction process and will also help in reducing the radiation related problems. (author)

  7. Effect of dissolved organic carbon in recycled wastewaters on boron adsorption by soils

    Science.gov (United States)

    In areas of water scarcity, recycled municipal wastewaters are being used as water resources for non-potable applications, especially for irrigation. Such wastewaters often contain elevated levels of dissolved organic carbon (DOC) and solution boron (B). Boron adsorption was investigated on eight ...

  8. Complexation with dissolved organic matter and solubility control of heavy metals in sandy soil

    NARCIS (Netherlands)

    Weng, L.; Temminghoff, E.J.M.; Lofts, S.; Tipping, E.; Riemsdijk, van W.H.

    2002-01-01

    The complexation of heavy metals with dissolved organic matter (DOM) in the environment influences the solubility and mobility of these metals. In this paper, we measured the complexation of Cu, Cd, Zn, Ni, and Pb with DOM in the soil solution at pH 3.7-6.1 using a Donnan membrane technique. The

  9. The use of curium neutrons to verify plutonium in spent fuel and reprocessing wastes

    International Nuclear Information System (INIS)

    Miura, N.

    1994-05-01

    For safeguards verification of spent fuel, leached hulls, and reprocessing wastes, it is necessary to determine the plutonium content in these items. We have evaluated the use of passive neutron multiplicity counting to determine the plutonium content directly and also to measure the 240 Pu/ 244 Cm ratio for the indirect verification of the plutonium. Neutron multiplicity counting of the singles, doubles, and triples neutrons has been evaluated for measuring 240 Pu, 244 Cm, and 252 Cf. We have proposed a method to establish the plutonium to curium ratio using the hybrid k-edge densitometer x-ray fluorescence instrument plus a neutron coincidence counter for the reprocessing dissolver solution. This report presents the concepts, experimental results, and error estimates for typical spent fuel applications

  10. The release of dissolved nutrients and metals from coastal sediments due to resuspension

    Science.gov (United States)

    Kalnejais, Linda H.; Martin, William R.; Bothner, Michael H.

    2010-01-01

    Coastal sediments in many regions are impacted by high levels of contaminants. Due to a combination of shallow water depths, waves, and currents, these sediments are subject to regular episodes of sediment resuspension. However, the influence of such disturbances on sediment chemistry and the release of solutes is poorly understood. The aim of this study is to quantify the release of dissolved metals (iron, manganese, silver, copper, and lead) and nutrients due to resuspension in Boston Harbor, Massachusetts, USA. Using a laboratory-based erosion chamber, a range of typical shear stresses was applied to fine-grained Harbor sediments and the solute concentration at each shear stress was measured. At low shear stress, below the erosion threshold, limited solutes were released. Beyond the erosion threshold, a release of all solutes, except lead, was observed and the concentrations increased with shear stress. The release was greater than could be accounted for by conservative mixing of porewaters into the overlying water, suggesting that sediment resuspension enhances the release of nutrients and metals to the dissolved phase. To address the long-term fate of resuspended particles, samples from the erosion chamber were maintained in suspension for 90. h. Over this time, 5-7% of the particulate copper and silver was released to the dissolved phase, while manganese was removed from solution. Thus resuspension releases solutes both during erosion events and over a longer timescale due to reactions of suspended particles in the water column. The magnitude of the annual solute release during erosion events was estimated by coupling the erosion chamber results with a record of bottom shear stresses simulated by a hydrodynamic model. The release of dissolved copper, lead, and phosphate due to resuspension is between 2% and 10% of the total (dissolved plus particulate phase) known inputs to Boston Harbor. Sediment resuspension is responsible for transferring a significant

  11. Development of an alternative process for recovery of uranium from rejected plates in the manufacture of MTR type fuel elements

    International Nuclear Information System (INIS)

    Flores Gonzalez, Jocelyn Natalia

    2011-01-01

    This work discusses the recovery of enriched uranium in U 235 , from fuel plates rejected during the fuel elements manufacturing process for the La Reina Nuclear Studies Center, RECH-1, CCHEN. The plates have an aluminum based alloy coating, AISI-SAE 6061, with U 3 Si 2 powder distributed evenly inside and dispersed in an aluminum matrix. The high cost of enriched uranium means that it must be recovered from plates rejected in the production process because of non-compliance with the plate specifications, and also because some of them undergo destructive testing, to measure the aluminum coating's thickness on each side of the plate. The thickness of the uranium nucleus is measured as well and the size of the defects on the ends of the plate such as 'dog bone' and 'fish tail', that is, for the purposes of quality control. The first step in the process is carried out by dissolving the aluminum in a hot solution of NaOH in order to release the uranium silicide powder that is insoluble in the soda. A second step involves dissolving the uranium silicide in a hot HNO 3 solution, followed by washing and filtering, and then extracting the SX and analyzing its behavior during this stage. During the process 98.9% of the uranium is recovered together with a solution that is enough for the SX process given the experiences that were carried out in the extraction stage

  12. SPES, Fuel Cycle Optimization for LWR

    International Nuclear Information System (INIS)

    1973-01-01

    1 - Nature of physical problem solved: Determination of optimal fuel cycle at equilibrium for a light water reactor taking into account batch size, fuel enrichment, de-rating, shutdown time, cost of replacement energy. 2 - Method of solution: Iterative method

  13. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  14. 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

  15. Behavior of silicon in nitric media. Application to uranium silicides fuels reprocessing

    International Nuclear Information System (INIS)

    Cheroux, L.

    2001-01-01

    Uranium silicides are used in some research reactors. Reprocessing them is a solution for their cycle end. A list of reprocessing scenarios has been set the most realistic being a nitric dissolution close to the classic spent fuel reprocessing. This uranium silicide fuel contains a lot of silicon and few things are known about polymerization of silicic acid in concentrated nitric acid. The study of this polymerization allows to point out the main parameters: acidity, temperature, silicon concentration. The presence of aluminum seems to speed up heavily the polymerization. It has been impossible to find an analytical technique smart and fast enough to characterize the first steps of silicic acid polymerization. However the action of silicic species on emulsions stabilization formed by mixing them with an organic phase containing TBP has been studied, Silicon slows down the phase separation by means of oligomeric species forming complex with TBP. The existence of these intermediate species is short and heating can avoid any stabilization. When non irradiated uranium silicide fuel is attacked by a nitric solution, aluminum and uranium are quickly dissolved whereas silicon mainly stands in solid state. That builds a gangue of hydrated silica around the uranium silicide particulates without preventing uranium dissolution. A small part of silicon passes into the solution and polymerize towards the highly poly-condensed forms, just 2% of initial silicon is still in molecular form at the end of the dissolution. A thermal treatment of the fuel element, by forming inter-metallic phases U-Al-Si, allows the whole silicon to pass into the solution and next to precipitate. The behavior of silicon in spent fuels should be between these two situations. (author)

  16. Dissolving Microneedle Patches for Dermal Vaccination.

    Science.gov (United States)

    Leone, M; Mönkäre, J; Bouwstra, J A; Kersten, G

    2017-11-01

    The dermal route is an attractive route for vaccine delivery due to the easy skin accessibility and a dense network of immune cells in the skin. The development of microneedles is crucial to take advantage of the skin immunization and simultaneously to overcome problems related to vaccination by conventional needles (e.g. pain, needle-stick injuries or needle re-use). This review focuses on dissolving microneedles that after penetration into the skin dissolve releasing the encapsulated antigen. The microneedle patch fabrication techniques and their challenges are discussed as well as the microneedle characterization methods and antigen stability aspects. The immunogenicity of antigens formulated in dissolving microneedles are addressed. Finally, the early clinical development is discussed.

  17. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  18. Proportioning of {sup 79}Se and {sup 126}Sn long life radionuclides in the fission products solutions coming from spent fuels processing; Dosage des radionucleides a vie longue {sup 79}Se et {sup 126}Sn dans les solutions de produits de fission issues du traitement des combustibles nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Comte, J

    2001-11-01

    The determination of radionuclides present in waste resulting from the nuclear fuel reprocessing is a request from the regulatory authorities to ensure an optimal management of the storage sites. Long-lived radionuclides (T{sub 1/2} > 30 years) are particularly concerned owing to the fact that their impact must be considered for the long term. Safety studies have established a list of long-lived radionuclides (LLRN) whose quantification is essential for the management of the disposal site. Among these, several are pure {beta} emitters, present at low concentration levels in complex matrices. Their determination, by radiochemical method or mass spectrometry, involves selective chemical separations from the others {beta}/{gamma} emitters and from the measurement interfering elements. The work undertaken in this thesis relates to the development of analytical methods for the determination of two long-lived radionuclides: selenium 79 and tin 126, in acid solutions of fission products present in nuclear fuel reprocessing plant. For selenium 79, a {beta} emitter with a half live estimated to be 10{sup 6} years, the bibliography describes different chemical separation methods including precipitation, liquid-liquid extraction and chromatography on ionic resins. After optimisation on a synthetic solution, two of these techniques, precipitation by potassium iodine and separation with ion exchange resins were applied to a genuine solution of fission products at Cogema La Hague. The results showed that only the ion exchange method allows us to obtain a solution sufficiently decontaminated (FD{beta}{gamma} = 250) with a significant selenium recovery yield (85%). This separation allows the measurement of the {sup 79}Se by electrothermal vaporization coupled with inductively coupled plasma mass spectrometry (ETV-ICP/MS), after transfer of the samples to CEA/Cadarache. The concentration of {sup 79}Se measured is 0,42 mg/L in the solution of fission products with an isotopic ratio

  19. Novel designs of continuous process for dissolution of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Rance, P.J.W.; Tinsley, T.P.; Polyakov, A.S.; Raginsky, L.S.; Morkovnikov, V.E.; Morozov, N.V.; Eliseev, S.P.

    1998-01-01

    A novel design of continuous dissolver for the dissolution of irradiated nuclear fuels is described. The development of the dissolver has resulted from a successful collaboration over the last four years between British Nuclear Fuels plc (UK) and the A.A. Bochvar All-Russia Research Institute of Inorganic Materials (Russia). An overview of the development work carried out on three different models is presented, and results from each of these are discussed. The dissolver provides many advantages over current designs of dissolvers. (author)

  20. Safety substantiation for underground isolation of spent nuclear fuel or spent nuclear materials as a basis to develop reliable technological solutions

    International Nuclear Information System (INIS)

    Gupalo, T.A.; Beygul, V.P.; Gupalo, M.S.; Kudinov, K.G.

    2000-01-01

    Major issues of the technique for mining and ecological safety substantiation of multi-barrier systems for long-term underground isolation of spent nuclear materials and solidified wastes containing long-lived radionuclides have been presented. The experience with the use of this technique for assessment of ecological safety for the long-term storage of plutonium-containing intermediate level wastes in underground facilities existing in the crystalline rock mass has been considered. The probabilistic evaluations of events of the emergency sequences of abnormal situations are based on the results of 40-year in-situ investigations in the rock mass. Feasibility of optimization has been shown for technological design solutions on storage facilities by the ''risk-costs'' principle. (authors)

  1. Determination of uranium in coated fuel particle compact by potassium fluoride fusion-gravimetric method

    International Nuclear Information System (INIS)

    Ito, Mitsuo; Iso, Shuichi; Hoshino, Akira; Suzuki, Shuichi.

    1992-03-01

    Potassium fluoride-gravimetric method has been developed for the determination of uranium in TRISO type-coated fuel particle compact. Graphite matrix in the fuel compact is burned off by heating it in a platinum crucible at 850degC. The coated fuel particles thus obtained are decomposed by fusion with potassium fluoride at 900degC. The melt was dissolved with sulfuric acid. Uranium is precipitated as ammonium diuranate, by passing ammonia gas through the solution. The resulting precipitate is heated in a muffle furnace at 850degC, to convert uranium into triuranium octoxide. Uranium in the triuranium octoxide was determined gravimetrically. Ten grams of caoted fuel particles were completely decomposed by fusion with 50 g of potassium fluoride at 900degC for 3 hrs. Analytical result for uranium in the fuel compact by the proposed method was 21.04 ± 0.05 g (n = 3), and was in good agreement with that obtained by non-destructive γ-ray measurement method : 21.01 ± 0.07 g (n = 3). (author)

  2. A New Control and Design of PEM Fuel Cell System Powered Diffused Air Aeration System

    Directory of Open Access Journals (Sweden)

    Hassen T. Dorrah

    2012-06-01

    Full Text Available The goal of aquaculture ponds is to maximize production and profits while holding labor and management efforts to the minimum. Poor water quality in most ponds causes risk of fish kills, disease outbreaks which lead to minimization of pond production. Dissolved Oxygen (DO is considered to be among the most important water quality parameters in fish culture. Fish ponds in aquaculture farms are usually located in remote areas where grid lines are at far distance. Aeration of ponds is required to prevent mortality and to intensify production, especially when feeding is practical, and in warm regions. To increase pond production it is necessary to control dissolved oxygen. Aeration offers the most immediate and practical solution to water quality problems encountered at higher stocking and feeding rates. Many units of aeration system are electrical units so using a continuous, high reliability, affordable, and environmentally friendly power sources is necessary. Fuel cells have become one of the major areas of research in the academia and the industry. Aeration of water by using PEM fuel cell power is not only a new application of the renewable energy, but also, it provides an affordable method to promote biodiversity in stagnant ponds and lakes. This paper presents a new design and control of PEM fuel cell powered a diffused air aeration system for a shrimp farm in Mersa Matruh in Egypt. Also Artificial intelligence (AI control techniques are used to control the fuel cell output power by controlling its input gases flow rate. Moreover the mathematical modeling and simulation of PEM fuel cell is introduced. A comparative study is applied between the performance of fuzzy logic controller (FLC and neural network controller (NNC. The results show the effectiveness of NNC over FLC.

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  4. 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.)

  5. XRF intermediate thickness layer technique for analysis of residue of hard to dissolve materials

    International Nuclear Information System (INIS)

    Mzyk, Z.; Mzyk, J.; Buzek, L.; Baranowska, I.

    1998-01-01

    This work presents a quick method for lead and silver determination in materials, such as slags from silver metallurgy and slimes from copper electrorefining, which are very difficult to dissolve, even using a microwave technique. The idea was to dissolve the possibly greatest amount of the sample using acids. Insoluble deposit was filtered out. Silver content in the solution was analysed by potentiometric titration or AAS, lead content by XRS, while sediment deposit on filter - by XRF intermediate thickness technique. The results of silver and lead analysis obtained by this method were compared with those obtained by classical method, i.e. melting the residue with sodium peroxide. (author)

  6. Automotive Fuel Processor Development and Demonstration with Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nuvera Fuel Cells

    2005-04-15

    The potential for fuel cell systems to improve energy efficiency and reduce emissions over conventional power systems has generated significant interest in fuel cell technologies. While fuel cells are being investigated for use in many applications such as stationary power generation and small portable devices, transportation applications present some unique challenges for fuel cell technology. Due to their lower operating temperature and non-brittle materials, most transportation work is focusing on fuel cells using proton exchange membrane (PEM) technology. Since PEM fuel cells are fueled by hydrogen, major obstacles to their widespread use are the lack of an available hydrogen fueling infrastructure and hydrogen's relatively low energy storage density, which leads to a much lower driving range than conventional vehicles. One potential solution to the hydrogen infrastructure and storage density issues is to convert a conventional fuel such as gasoline into hydrogen onboard the vehicle using a fuel processor. Figure 2 shows that gasoline stores roughly 7 times more energy per volume than pressurized hydrogen gas at 700 bar and 4 times more than liquid hydrogen. If integrated properly, the fuel processor/fuel cell system would also be more efficient than traditional engines and would give a fuel economy benefit while hydrogen storage and distribution issues are being investigated. Widespread implementation of fuel processor/fuel cell systems requires improvements in several aspects of the technology, including size, startup time, transient response time, and cost. In addition, the ability to operate on a number of hydrocarbon fuels that are available through the existing infrastructure is a key enabler for commercializing these systems. In this program, Nuvera Fuel Cells collaborated with the Department of Energy (DOE) to develop efficient, low-emission, multi-fuel processors for transportation applications. Nuvera's focus was on (1) developing fuel

  7. CADDIS Volume 2. Sources, Stressors and Responses: Dissolved Oxygen - Detailed Conceptual Diagram

    Science.gov (United States)

    Introduction to the dissolved oxygen module, when to list dissolved oxygen as a candidate cause, ways to measure dissolved oxygen, simple and detailed conceptual model diagrams for dissolved oxygen, references for the dissolved oxygen module.

  8. CADDIS Volume 2. Sources, Stressors and Responses: Dissolved Oxygen - Simple Conceptual Diagram

    Science.gov (United States)

    Introduction to the dissolved oxygen module, when to list dissolved oxygen as a candidate cause, ways to measure dissolved oxygen, simple and detailed conceptual model diagrams for dissolved oxygen, references for the dissolved oxygen module.

  9. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  10. Advanced Nuclear Fuels for More Capable and Sustainable Exploration

    Data.gov (United States)

    National Aeronautics and Space Administration — Molten salt reactors are a subtype of reactor that uses nuclear fuel dissolved in a molten salt liquid medium (such as LiF-BeF2-UF4) as both fuel and coolant. The...

  11. Fabrication of simulated DUPIC fuel

    Science.gov (United States)

    Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

    2000-12-01

    Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

  12. 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.)

  13. Electro-volatilization of ruthenium in nitric medium: influences of ruthenium species nature and models solutions composition

    International Nuclear Information System (INIS)

    Mousset, F.

    2004-12-01

    Ruthenium is one of the fission products in the reprocessing of irradiated fuels that requires a specific processing management. Its elimination, upstream by the PUREX process, has been considered. A process, called electro-volatilization, which take advantage of the RuO 4 volatility, has been optimised in the present study. It consists in a continuous electrolysis of ruthenium solutions in order to generate RuO 4 species that is volatilized and easily trapped. This process goes to satisfying ruthenium elimination yields with RuNO(NO 3 ) 3 (H 2 O) 2 synthetic solutions but not with fuel dissolution solutions. Consequently, this work consisted in the speciation studies of dissolved ruthenium species were carried out by simulating fuel solutions produced by hot acid attack of several ruthenium compounds (Ru(0), RuO 2 ,xH 2 O, polymetallic alloy). In parallel with dissolution kinetic studies, the determination of dissolved species was performed using voltammetry, spectrometry and spectro-electrochemistry. The results showed the co-existence of Ru(IV) and RuNO(NO 2 ) 2 (H 2 O) 3 . Although these species are different from synthetic RuNO(NO 3 ) 3 (H 2 O) 2 , their electro-oxidation behaviour are similar. The electro-volatilization tests of these dissolution solutions yielded to comparable results as the synthetic RuNO(NO 3 ) 3 (H 2 O) 2 solutions. Then, complexity increase of models solutions was performed by in-situ generation of nitrous acid during ruthenium dissolution. Nitrous acid showed a catalytic effect on ruthenium dissolution. Its presence goes to quasi exclusively RuNO(NO 2 ) 2 (H 2 O) 3 species. It is also responsible of the strong n-bond formation between Ru 2+ and NO + . In addition, it has been shown that its reducing action on RuO 4 hinders the electro-volatilization process. Mn 2+ and Ce 3+ cations also reveal, but to a lesser extent, an electro-eater behaviour as well as Pu 4+ and Cr 3+ according to the thermodynamics data. These results allow one to

  14. Dissolved petroleum hydrocarbons in the Andaman Sea

    Digital Repository Service at National Institute of Oceanography (India)

    Topgi, R.S.; Noronha, R.J.; Fondekar, S.P.

    Mean dissolved petroleum hydrocarbons, measured using UV-spectrophotometry, at 0 and 10m were 51 plus or minus 1 and 55 plus or minus 1.2 mu g/litre respectively; range of variation being between 28 and 83 mu g/litre. Very little difference...

  15. Dissolved carbohydrate in the central Arabian Sea

    Digital Repository Service at National Institute of Oceanography (India)

    Dhople, V.M.; Bhosle, N.B.

    with chlorophyll a (P 0.001) and phaeopigments (P 0.001) suggesting its release from the former and zooplankton grazing in the latter. Inverse correlations with dissolved oxygen, phosphate and nitrate indicated the possibility of the release of carbohydrate from...

  16. Modeling Fish Growth in Low Dissolved Oxygen

    Science.gov (United States)

    Neilan, Rachael Miller

    2013-01-01

    This article describes a computational project designed for undergraduate students as an introduction to mathematical modeling. Students use an ordinary differential equation to describe fish weight and assume the instantaneous growth rate depends on the concentration of dissolved oxygen. Published laboratory experiments suggest that continuous…

  17. Total dissolved carbohydrate in Mahi river estuary

    Digital Repository Service at National Institute of Oceanography (India)

    Bhosle, N.B.; Rokade, M.A.; Zingde, M.D.

    Total dissolved carbohydrate varied from 4.37-15 mg l-1 and 3.71-15.95 mg l-1 in the surface and bottom samples respectively. Highest concentration of carbohydrate was observed at station 1 which decreased downward upto Station 6 which showed...

  18. Release of dissolved 85Kr by standing

    International Nuclear Information System (INIS)

    Ootsuka, Norikatsu; Yamamoto, Tadatoshi; Tsukui, Kohei

    1986-01-01

    The experiments on the release of dissolved 85 Kr by standing at room temperature were carried out to examine the influence of liquid level in a sampler and properties of solvent on the release efficiency. Six kinds of organic solvents as well as water were taken as solvents. The half-life period in case of the decrease in concentration of the dissolved 85 Kr which was used as an index of release efficiency, was proportional to the liquid level in the sampler and was inversely proportional to the diffusion coefficient of Kr gas in solvent. For organic solvents belonging to homologous series, the half-life period became longer with increasing the carbon number of solvent molecule. From the relationship between the half-life period and the carbon number, the release efficiency in the dissolved 85 Kr can be predicted for any commonly used solvent as a practical application. This method was found to be an effective means of removing the dissolved 85 Kr of low level though it takes rather long time. (author)

  19. Subcooled boiling effect on dissolved gases behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.; Sinkule, J.; Linek, V.

    1999-01-01

    A model describing dissolved gasses (hydrogen, nitrogen) and ammonia behaviour in subcooled boiling conditions of WWERs was developed. Main objective of the study was to analyse conditions and mechanisms leading to formation of a zone with different concentration of dissolved gases, eg. a zone depleted in dissolved hydrogen in relation to the bulk of coolant. Both, an equilibrium and dynamic approaches were used to describe a depletion of the liquid surrounding a steam bubble in the gas components. The obtained results show that locally different water chemistry conditions can be met in the subcooled boiling conditions, especially, in the developed subcooled boiling regime. For example, a 70% hydrogen depletion in relation to the bulk of coolant takes about 1 ms and concerns a liquid layer of 1 μn surrounding the steam bubble. The locally different concentration of dissolved gases can influence physic-chemical and radiolytic processes in the reactor system, eg. Zr cladding corrosion, radioactivity transport and determination of the critical hydrogen concentration. (author)

  20. Radiocarbon in marine dissolved organic carbon (DOC)

    NARCIS (Netherlands)

    Clercq, M. le; Plicht, J. van der; Meijer, H.A.J.; Baar, H.J.W. de

    Dissolved Organic Carbon (DOC) plays an important role in the ecology and carbon cycle in the ocean. Analytical problems with concentration and isotope ratio measurements have hindered its study. We have constructed a new analytical method based on supercritical oxidation for the determination of

  1. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  2. The importance of dissolved free oxygen during formation of sandstone-type uranium deposits

    Science.gov (United States)

    Granger, Harry Clifford; Warren, C.G.

    1979-01-01

    One factor which distinguishes t, he genesis of roll-type uranium deposits from the Uravan Mineral Belt and other sandstone-type uranium deposits may be the presence and concentration of dissolved free oxygen in the ore-forming. solutions. Although dissolved oxygen is a necessary prerequisite for the formation of roll-type deposits, it is proposed that a lack of dissolved oxygen is a prerequisite for the Uravan deposits. Solutions that formed both types of deposits probably had a supergene origin and originated as meteoric water in approximate equilibrium with atmospheric oxygen. Roll-type deposits were formed where the Eh dropped abruptly following consumption of the oxygen by iron sulfide minerals and creation of kinetically active sulfur species that could reduce uranium. The solutions that formed the Uravan deposits, on the other hand, probably first equilibrated with sulfide-free ferrous-ferric detrital minerals and fossil organic matter in the host rock. That is, the uraniferous solutions lost their oxygen without lowering their Eh enough to precipitate uranium. Without oxygen, they then. became incapable of oxidizing iron sulfide minerals. Subsequent localization and formation of ore bodies from these oxygen-depleted solutions, therefore, was not necessarily dependent on large reducing capacities.

  3. Removal of actinides from dissolved ORNL MVST sludge using the TRUEX process

    International Nuclear Information System (INIS)

    Spencer, B.B.; Egan, B.Z.; Chase, C.W.

    1997-07-01

    Experiments were conducted to evaluate the transuranium extraction process for partitioning actinides from actual dissolved high-level radioactive waste sludge. All tests were performed at ambient temperature. Time and budget constraints permitted only two experimental campaigns. Samples of sludge from Melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign, the rinsed sludge was dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 1.8 M with a nitric acid concentration of ca. 2.9 M. About 50% of the dry mass of the sludge was dissolved. In the other campaign, the sludge was neutralized with nitric acid to destroy the carbonates, then leached with ca. 2.6 M NaOH for ca. 6 h before rinsing with the mild caustic. The sludge was then dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 0.6 M with a nitric acid concentration of ca. 1.7 M. About 80% of the sludge dissolved. The dissolved sludge solution form the first campaign began gelling immediately, and a visible gel layer was observed after 8 days. In the second campaign, the solution became hazy after ca. 8 days, indicating gel formation, but did not display separated gel layers after aging for 20 days. Batch liquid-liquid equilibrium tests of both the extraction and stripping operations were conducted. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th, and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  5. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  6. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gonzalez, J.L.

    2002-01-01

    The spent fuel management strategy in Spain is presented. The strategy includes temporary solutions and plans for final disposal. The need for R and D including partitioning and transmutation, as well as the financial constraints are also addressed. (author)

  7. The impact of UV irradiation on the radical initiating capacity of dissolved dyes

    International Nuclear Information System (INIS)

    Vig, A.; Czilik, M.; Rusznak, I.

    2002-01-01

    Complete text of publication follows. Kinetics of photodecomposition of three model dyes dissolved in isopropanol-water mixture has been determined after exposure to UV radiation in the range from 360 through 400 nm and from 220 through 400 nm, respectively. It has been disclosed earlier that photodecomposition of the dissolved dyes was decelerated initially by the presence of the dissolved oxygen in the system. The presence of a radical initiator, AIBN was indispensable for arriving at the decomposition of the irradiated dye solution in the range from 360 through 400 nm. The equation of W i D = [O 2 ]/τ D was used for the calculation of radical initiating rate of the irradiated dye molecule on the isopropanol (W i D (mol/l x s)), where [O 2 ] (mol/l) is the dissolved oxygen concentration in the system and τ D (s) is duration of the induction period of the photodestruction of the dissolved dye. The equation is valid only for photodecomposition which are not chain reaction. The photodegradation of dissolved dyes was also other then chain reaction, consequently the above equation could be applied in the study too. The average radical initiating rate of the dyes applied in this study was in the order of magnitude equal to that of AIBN. The number of cycles between the first radical formation and the last regeneration of the dye molecule could be calculated in bath systems (in the presence and absence of oxygen, respectively): K = W i D /W D , where K is the number of cycles, W D (mol/l x s) is the initial rate of the decomposition of the dissolved dyed. The number of cycles in the oxygen containing systems significantly exceeded those obtained in the oxygen systems because W D was markedly higher in the latter system than in the former one

  8. Basin-scale transport of hydrothermal dissolved metals across the South Pacific Ocean.

    Science.gov (United States)

    Resing, Joseph A; Sedwick, Peter N; German, Christopher R; Jenkins, William J; Moffett, James W; Sohst, Bettina M; Tagliabue, Alessandro

    2015-07-09

    Hydrothermal venting along mid-ocean ridges exerts an important control on the chemical composition of sea water by serving as a major source or sink for a number of trace elements in the ocean. Of these, iron has received considerable attention because of its role as an essential and often limiting nutrient for primary production in regions of the ocean that are of critical importance for the global carbon cycle. It has been thought that most of the dissolved iron discharged by hydrothermal vents is lost from solution close to ridge-axis sources and is thus of limited importance for ocean biogeochemistry. This long-standing view is challenged by recent studies which suggest that stabilization of hydrothermal dissolved iron may facilitate its long-range oceanic transport. Such transport has been subsequently inferred from spatially limited oceanographic observations. Here we report data from the US GEOTRACES Eastern Pacific Zonal Transect (EPZT) that demonstrate lateral transport of hydrothermal dissolved iron, manganese, and aluminium from the southern East Pacific Rise (SEPR) several thousand kilometres westward across the South Pacific Ocean. Dissolved iron exhibits nearly conservative (that is, no loss from solution during transport and mixing) behaviour in this hydrothermal plume, implying a greater longevity in the deep ocean than previously assumed. Based on our observations, we estimate a global hydrothermal dissolved iron input of three to four gigamoles per year to the ocean interior, which is more than fourfold higher than previous estimates. Complementary simulations with a global-scale ocean biogeochemical model suggest that the observed transport of hydrothermal dissolved iron requires some means of physicochemical stabilization and indicate that hydrothermally derived iron sustains a large fraction of Southern Ocean export production.

  9. Electrodialysis-ion exchange for the separation of dissolved salts

    International Nuclear Information System (INIS)

    Baroch, C.J.; Grant, P.J.

    1995-01-01

    The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. This report describes the process of electrodialysis-ion exchange (EDIX) for treating aqueous wastes streams consisting of nitrates, sodium, organics, heavy metals, and radioactive species

  10. Electrodialysis-ion exchange for the separation of dissolved salts

    Energy Technology Data Exchange (ETDEWEB)

    Baroch, C.J. [Wastren, Inc., Westminster, CO (United States); Grant, P.J. [Wastren, Inc., Hummelstown, PA (United States)

    1995-10-01

    The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. There is considerable interest in developing processes that remove or destroy the nitrate wastes. Electrodialysis-Ion Exchange (EDIX) is a possible process that should be more cost effective in treating aqueous waste steams. This report describes the EDIX process.

  11. Microwave irradiated Ni–MnO{sub x}/C as an electrocatalyst for methanol oxidation in KOH solution for fuel cell application

    Energy Technology Data Exchange (ETDEWEB)

    Hameed, R.M. Abdel, E-mail: randa311eg@yahoo.com

    2015-12-01

    Graphical abstract: - Highlights: • Ni–MnO{sub x}/C had nickel nanoparticles with an average diameter of 4.5 nm. • Oxidation current density increased by 1.43 times at Ni–MnO{sub x}/C. • Ni–MnO{sub x}/C showed k{sub s} value of 3.26 × 103 cm{sup 3} mol{sup −1} s{sup −1}. • Adding MnO{sub x} to Ni/C lowered its phase angle and impedance values. - Abstract: Ni–MnO{sub x}/C electrocatalyst was synthesized by the reduction of nickel precursor salt on MnO{sub x}/C powder using NaBH{sub 4} and the deposition process was motivated with the aid of microwave irradiation. Finer nickel nanoparticles were detected in Ni–MnO{sub x}/C using transmission electron microscopy with a lower particle size of 4.5 nm compared to 6 nm in Ni/C. Cyclic voltammetry, chronoamperometry and electrochemical impedance spectroscopy (EIS) were applied to study the electrocatalytic activity of Ni–MnO{sub x}/C for methanol oxidation in 0.5 M KOH solution. The presence of 7.5 wt.% MnO{sub x} in Ni–MnO{sub x}/C enhanced the oxidation current density by 1.43 times. The catalytic rate constant of methanol oxidation at Ni–MnO{sub x}/C was calculated as 3.26 × 10{sup 3} cm{sup 3} mol{sup −1} s{sup −1}. An appreciable shift in the maximum frequency at the transition from the resistive to capacitive regions to a higher value in Bode plots of Ni–MnO{sub x}/C was shown when compared to Ni/C. It was accompanied by lowered phase angle values. The lowered Warburg impedance value (W) of Ni–MnO{sub x}/C at 400 mV confirmed the faster methanol diffusion rate at its surface.

  12. Coulometric determination of dissolved hydrogen with a multielectrolytic modified carbon felt electrode-based sensor.

    Science.gov (United States)

    Matsuura, Hiroaki; Yamawaki, Yosuke; Sasaki, Kosuke; Uchiyama, Shunichi

    2013-06-01

    A multielectrolytic modified carbon electrode (MEMCE) was fabricated by the electrolytic-oxidation/reduction processes. First, the functional groups containing nitrogen atoms such as amino group were introduced by the electrode oxidation of carbon felt electrode in an ammonium carbamate aqueous solution, and next, this electrode was electroreduced in sulfuric acid. The redox waves between hydrogen ion and hydrogen molecule at highly positive potential range appeared in the cyclic voltammogram obtained by MEMCE. A coulometric cell using MEMCE with a catalytic activity of electrooxidation of hydrogen molecule was constructed and was used for the measurement of dissolved hydrogen. The typical current vs. time curve was obtained by the repetitive measurement of the dissolved hydrogen. These curves indicated that the measurement of dissolved hydrogen was finished completely in a very short time (ca. 10 sec). A linear relationship was obtained between the electrical charge needed for the electrooxidation process of hydrogen molecule and dissolved hydrogen concentration. This indicates that the developed coulometric method can be used for the determination of the dissolved hydrogen concentration.

  13. Pulsating potentiometric titration technique for assay of dissolved oxygen in water at trace level.

    Science.gov (United States)

    Sahoo, P; Ananthanarayanan, R; Malathi, N; Rajiniganth, M P; Murali, N; Swaminathan, P

    2010-06-11

    A simple but high performance potentiometric titration technique using pulsating sensors has been developed for assay of dissolved oxygen (DO) in water samples down to 10.0 microg L(-1) levels. The technique involves Winkler titration chemistry, commonly used for determination of dissolved oxygen in water at mg L(-1) levels, with modification in methodology for accurate detection of end point even at 10.0 microg L(-1) levels DO present in the sample. An indigenously built sampling cum pretreatment vessel has been deployed for collection and chemical fixing of dissolved oxygen in water samples from flowing water line without exposure to air. A potentiometric titration facility using pulsating sensors developed in-house is used to carry out titration. The power of the titration technique has been realised in estimation of very dilute solution of iodine equivalent to 10 microg L(-1) O(2). Finally, several water samples containing dissolved oxygen from mg L(-1) to microg L(-1) levels were successfully analysed with excellent reproducibility using this new technique. The precision in measurement of DO in water at 10 microg L(-1) O(2) level is 0.14 (n=5), RSD: 1.4%. Probably for the first time a potentiometric titration technique has been successfully deployed for assay of dissolved oxygen in water samples at 10 microg L(-1) levels. Copyright 2010 Elsevier B.V. All rights reserved.

  14. Pulsating potentiometric titration technique for assay of dissolved oxygen in water at trace level

    International Nuclear Information System (INIS)

    Sahoo, P.; Ananthanarayanan, R.; Malathi, N.; Rajiniganth, M.P.; Murali, N.; Swaminathan, P.

    2010-01-01

    A simple but high performance potentiometric titration technique using pulsating sensors has been developed for assay of dissolved oxygen (DO) in water samples down to 10.0 μg L -1 levels. The technique involves Winkler titration chemistry, commonly used for determination of dissolved oxygen in water at mg L -1 levels, with modification in methodology for accurate detection of end point even at 10.0 μg L -1 levels DO present in the sample. An indigenously built sampling cum pretreatment vessel has been deployed for collection and chemical fixing of dissolved oxygen in water samples from flowing water line without exposure to air. A potentiometric titration facility using pulsating sensors developed in-house is used to carry out titration. The power of the titration technique has been realised in estimation of very dilute solution of iodine equivalent to 10 μg L -1 O 2 . Finally, several water samples containing dissolved oxygen from mg L -1 to μg L -1 levels were successfully analysed with excellent reproducibility using this new technique. The precision in measurement of DO in water at 10 μg L -1 O 2 level is 0.14 (n = 5), RSD: 1.4%. Probably for the first time a potentiometric titration technique has been successfully deployed for assay of dissolved oxygen in water samples at 10 μg L -1 levels.

  15. Origins and bioavailability of dissolved organic matter in groundwater

    Science.gov (United States)

    Shen, Yuan; Chapelle, Francis H.; Strom, Eric W.; Benner, Ronald

    2015-01-01

    Dissolved organic matter (DOM) in groundwater influences water quality and fuels microbial metabolism, but its origins, bioavailability and chemical composition are poorly understood. The origins and concentrations of dissolved organic carbon (DOC) and bioavailable DOM were monitored during a long-term (2-year) study of groundwater in a fractured-rock aquifer in the Carolina slate belt. Surface precipitation was significantly correlated with groundwater concentrations of DOC, bioavailable DOM and chromophoric DOM, indicating strong hydrological connections between surface and ground waters. The physicochemical and biological processes shaping the concentrations and compositions of DOM during its passage through the soil column to the saturated zone are conceptualized in the regional chromatography model. The model provides a framework for linking hydrology with the processes affecting the transformation, remineralization and microbial production of DOM during passage through the soil column. Lignin-derived phenols were relatively depleted in groundwater DOM indicating substantial removal in the unsaturated zone, and optical properties of chromophoric DOM indicated lower molecular weight DOM in groundwater relative to surface water. The prevalence of glycine, γ-aminobutyric acid, and d-enantiomers of amino acids indicated the DOM was highly diagenetically altered. Bioassay experiments were used to establish DOC-normalized yields of amino acids as molecular indicators of DOM bioavailability in groundwater. A relatively small fraction (8 ± 4 %) of DOC in groundwater was bioavailable. The relatively high yields of specific d-enantiomers of amino acids indicated a substantial fraction (15–34 %) of groundwater DOC was of bacterial origin.

  16. Study on the Electrochemical Behavior of Iodide at Platinum Electrode in Potassium Chlorate Solution

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Sang Hyuk; Yeon, Jei Won; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Radioactive iodine-131, is one of the most hazardous fission products which could be released from fuels of nuclear reactors during the severe accident of nuclear power plants. Due to its high radioactivity, high fission yield (2.8%) and hazardous biological effects, the behavior of iodine has been taken interests in many research groups. Iodine is known to be released from the fuels as a cesium iodide form, CsI. And, as nuclear fuels are mostly placed in the water pool, it is easily dissolved in the water after released from the fuels. In water, iodide anion could be oxidized into molecular iodine. As the molecular iodine is a volatile species and the oxidizing rate is affected by many environmental facts such as pH, radiolysis products and temperature, the oxidation reaction of the iodide ion has been considered as an important chemical reaction related to the severe accident of nuclear power plants In present work, the electrochemical behavior of iodide anion was observed by using cyclic voltammetric technique in potassium chlorate solutions. We observed two different oxidation waves in the oxidation potential region. From the comparison with the previous reported results, one is regarded as the oxidation of iodide into molecular iodine. The other is evaluated to be the formation of high-valent iodine-containing compounds

  17. Dissolved oxygen detection by galvanic displacement-induced

    Indian Academy of Sciences (India)

    Dissolved oxygen detection by galvanic displacement-induced graphene/silver nanocomposite ... dissolved oxygen (DO) detection based on a galvanic displacement synthesized reduced graphene oxide–silver nanoparticles ... Current Issue

  18. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding