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Sample records for fuel dissolver solution

  1. 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.)

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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)

  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. 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)

  9. 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.)

  10. 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

  11. 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)

  12. 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)

  13. 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.

  14. 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.)

  15. 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)

  16. 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

  17. 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.)

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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)

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  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. 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

  10. 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.

  11. 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.

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. 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)

  2. 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

  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. 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.)

  5. 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

  6. 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)

  7. 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)

  8. 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)

  9. 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)

  10. 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

  11. 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

  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 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.

  14. 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.)

  15. 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

  16. 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)

  17. 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)

  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. 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.

  20. 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

  1. 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.

  2. 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?

  3. 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.)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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

  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. 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

  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. 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)

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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.

  10. 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.

  11. 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)

  12. 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...

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. 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

  19. 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.)

  20. 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.

  1. 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

  2. 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)

  3. 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.

  4. [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.

  5. 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

  6. 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)

  7. 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.

  8. [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.

  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. 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.)

  11. 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.)

  12. 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).

  13. 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

  14. 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

  15. 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)

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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...

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. 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.

  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. 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.

  15. 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

  16. [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.

  17. 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

  18. 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.

  19. 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)

  20. 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.

  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. 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

  3. 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.)

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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)

  9. 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...

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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)

  14. DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

    Science.gov (United States)

    Swanson, J.L.

    1961-07-11

    The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

  15. 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

  16. 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

  17. 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%

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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.)

  14. 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

  15. 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

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.)

  3. 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

  4. 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

  5. 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.

  6. Uranium chloride extraction of transuranium elements from LWR fuel

    International Nuclear Information System (INIS)

    Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure

  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. Magnesium transport extraction of transuranium elements from LWR fuel

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E.; Pierce, R.D.

    1992-01-01

    This patent describes a process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuel containing rare earth and noble metal fission products as well as fission products of alkali metals, alkaline earth metals and iodine. It comprises reducing the oxide fuel with Ca metal in the presence of Ca halide; separating the Ca halide with the CaO and the fission products contained therein from the U-Fe alloy and the metal values dissolved therein and electrolytically contacting the calcium salts with a carbon electrode; contacting the liquid U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel with liquid Mg metal, thereafter separating the liquid Mg and the metals dissolved therein from the U-Fe alloy and the metal dissolved therein, distilling the Mg from the transuranium actinide and rare earth metals, recontacting the U-Fe alloy with liquid Mg metal a sufficient number of times until not less than about 99% by weight of the transuranium actinide values have been removed from the U-Fe alloy

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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.

  17. Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shehee, T. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jones, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); DelCul, G. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-10-02

    The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pu and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.

  18. 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)

  19. 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

  20. 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.

  1. 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

  2. Analysis of iodine chemical form noted from severe fuel damage experiments

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Osetek, D.J.

    1986-01-01

    Data from the TMI-2 accident has shown that only small amounts of iodine (I) escaped the plant. The postulated reason for such limited release is the formation of CsI (a salt) within fuel, which remains stable in a reducing high-temperature steam-H 2 environment. Upon cooldown CsI would dissolve in water condensate to form an ionic solution. However, recent data from fuel destruction experiments indicate different iodine release behavior that is tied to fuel burnup and oxidation conditions, as well as fission product concentration levels in the steam/H 2 effluent. Analysis of the data indicate that at low-burnup conditions, atomic I release from fuel is favored. Likewise, at low fission product concentration conditions HI is the favored chemical form in the steam/H 2 environment, not CsI. Results of thermochemical equilibria and chemical kinetics analysis support the data trends noted from the PBF-SFD tests. An a priori assumption of CsI for risk analysis of all accident sequences may therefore be inappropriate

  3. Feasibility study of palm-based fuels for hybrid rocket motor applications

    Science.gov (United States)

    Tarmizi Ahmad, M.; Abidin, Razali; Taha, A. Latif; Anudip, Amzaryi

    2018-02-01

    This paper describes the combined analysis done in pure palm-based wax that can be used as solid fuel in a hybrid rocket engine. The measurement of pure palm wax calorific value was performed using a bomb calorimeter. An experimental rocket engine and static test stand facility were established. After initial measurement and calibration, repeated procedures were performed. Instrumentation supplies carried out allow fuel regression rate measurements, oxidizer mass flow rates and stearic acid rocket motors measurements. Similar tests are also carried out with stearate acid (from palm oil by-products) dissolved with nitrocellulose and bee solution. Calculated data and experiments show that rates and regression thrust can be achieved even in pure-tested palm-based wax. Additionally, palm-based wax is mixed with beeswax characterized by higher nominal melting temperatures to increase moisturizing points to higher temperatures without affecting regression rate values. Calorie measurements and ballistic experiments were performed on this new fuel formulation. This new formulation promises driving applications in a wide range of temperatures.

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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.)

  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. 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

  13. 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.)

  14. Electro-volatilization of ruthenium in nitric medium: influences of ruthenium species nature and models solutions composition; Electro-volatilisation du ruthenium en milieu nitrique: influence de la nature des formes chimiques du ruthenium et de la composition des solutions modeles de dissolution

    Energy Technology Data Exchange (ETDEWEB)

    Mousset, F

    2004-12-15

    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{sub 4} volatility, has been optimised in the present study. It consists in a continuous electrolysis of ruthenium solutions in order to generate RuO{sub 4} species that is volatilized and easily trapped. This process goes to satisfying ruthenium elimination yields with RuNO(NO{sub 3}){sub 3}(H{sub 2}O){sub 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{sub 2},xH{sub 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{sub 2}){sub 2}(H{sub 2}O){sub 3}. Although these species are different from synthetic RuNO(NO{sub 3}){sub 3}(H{sub 2}O){sub 2}, their electro-oxidation behaviour are similar. The electro-volatilization tests of these dissolution solutions yielded to comparable results as the synthetic RuNO(NO{sub 3}){sub 3}(H{sub 2}O){sub 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{sub 2}){sub 2}(H{sub 2}O){sub 3} species. It is also responsible of the strong n-bond formation between Ru{sup 2+} and NO{sup +}. In addition, it has been shown that its reducing action on RuO{sub 4} hinders the electro-volatilization process. Mn{sup 2+} and Ce{sup 3+} cations also reveal, but to a lesser

  15. 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.

  16. 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)

  17. 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

  18. 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)

  19. 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.

  20. 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

  1. 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...

  2. TMI-2 fuel-recovery plant. Feasibility study

    International Nuclear Information System (INIS)

    Evans, D.L.

    1982-12-01

    This project is a feasibility study for constructing a TMI-2 core Fuel Recovery Plant at the Idaho National Engineering Laboratory (INEL). The primary objectives of the Fuel Recovery Plant (FRP) are to recover and account for the fuel and to process, isolate, and package the waste material from the TMI-2 core. This feasibility study is predicated on a baseline plant and covers its design, fabrication, installation, testing and operation. Alternative methods for the disposal of the TMI-2 core have also been considered, but not examined in detail for their feasibility. The FRP will receive TMI-2 fuel in canisters. The fuel will vary from core debris to intact fuel assemblies and include some core structural materials. The canister contents will be shredded and subsequently fed to a dissolver. Uranium, plutonium, fission products, and some core structural material will be dissolved. The uranium will be separated by solvent extraction and solidified by calcination. The plutonium will also be separated by solvent extraction and routed to the Plutonium Extraction Facility. The wastes will be packaged for further treatment, temporary storage or permanent disposal

  3. The refurbishment of the D1206 fuel reprocessing plant

    International Nuclear Information System (INIS)

    Bailey, G.

    1988-01-01

    The term decommissioning can be applied not only to reactors but to any nuclear plant, laboratory, building or part of a building that may have been associated with radioactive material and needs to be restored to clean conditions. In this case the decommissioning and reconstruction of the Dounreay Fast Reactor fuel reprocessing plant, so that plutonium oxide could be reprocessed as well as enriched uranium fuel, is described. The work included improving containment and shielding, building a new head-end treatment cave for the more complex and larger fuel elements, improving the ventilation and constructing a new dissolver. In this paper the breakdown cave and dissolver cell are described and compared and the work done explained. (U.K.)

  4. 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

  5. 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

  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. National security

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    An x-ray-fluorescence analysis system is being developed for determining uranium and plutonium concentrations in the solution obtained when the fuel rods from light-water power reactors are dissolved in nitric acid. These concentrations are needed to inventory the uranium and plutonium handled in fuel reprocessing plants. This technique should make it possible to analyze dissolver solution with speed and a 0.5 percent accuracy

  8. 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

  9. Uranium and plutonium determinations for evaluation of high burnup fuel performance

    International Nuclear Information System (INIS)

    Heinrich, R.R.; Popek, R.J.; Bowers, D.L.; Essling, A.M.; Callis, E.L.; Persiani, P.J.

    1985-01-01

    Purpose of this work is to experimentally test computational methods being developed for reactor fuel operation. Described are the analytical techniques used in the determination of uranium and plutonium compositions on PWR fuel that has spanned five power cycles, culminating in 55,000 to 57,000 MWd/T burnup. Analyses have been performed on ten samples excised from selected sections of the fuel rods. Hot cell operations required the separation of fuel from cladding and the comminution of the fuel. These tasks were successfully accomplished using a SpectroMil, a ball pestle impact grinding and blending instrument manufactured by Chemplex Industries, Inc., Eastchester, New York. The fuel was dissolved using strong mineral acids and bomb dissolution techniques. Separation of the fuel from fission products was done by solvent (hexone) extraction. Fuel isotopic compositions and assays were determined by the mass spectrometric isotope dilution (MSID) method using NBS standards SRM-993 and SRM-996. Alpha spectrometry was used to determine the 238 Pu composition. Relative correlations of composition with burnup were obtained by gamma-ray spectrometry of selected fission products in the dissolved fuel

  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. 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

  12. 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.

  13. 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.

  14. Solution chemistry and separation of metal ions in leached solution

    International Nuclear Information System (INIS)

    Shibata, J.

    1991-01-01

    The method to presume a dissolved state of metal ions in an aqueous solution and the technology to separate and concentrate metal ions in a leached solution are described in this paper. It is very important for the separation of metal ions to know the dissolved state of metal ions. If we know the composition of an aqueous solution and the stability constants of metal-ligand complexes, we can calculate and estimate the concentration of each species in the solution. Then, we can decide the policy to separate and concentrate metal ions. There are several methods for separation and purification; hydroxide precipitation method, sulfide precipitation method, solvent extraction method and ion exchange resin method. Solvent extraction has been used in purification processes of copper refinery, uranium refinery, platinum metal refinery and rare earth metal refinery. Fundamental process of solvent extraction, a kind of commercial extractants, a way of determining a suitable extractant and an equipment are discussed. Finally, it will be emphasized how the separation of rare earths is improved in solvent extraction. (author) 21 figs., 8 tabs., 8 refs

  15. Comparison of uranium dissolution rates from spent fuel and uranium dioxide

    International Nuclear Information System (INIS)

    Steward, S.A.; Gray, W.J.

    1994-01-01

    Two similar sets of dissolution experiments, resulting from a statistical experimental design were performed in order to examine systematically the effects of temperature (25--75 degree C), dissolved oxygen (0.002-0.2 atm overpressure), pH (8--10) and carbonate concentrations (2--200 x 10 -4 molar) on aqueous dissolution of UO 2 and spent fuel. The average dissolution rate was 8.6 mg/m 2 ·day for UO 2 and 3.1 mg/m 2 ·day for spent fuel. This is considered to be an insignificant difference; thus, unirradiated UO 2 and irradiated spent fuel dissolved at about the same rate. Moreover, regression analyses indicated that the dissolution rates of UO 2 and spent fuel responded similarly to changes in pH, temperature, and carbonate concentration. However, the two materials responded very differently to dissolved oxygen concentration. Approximately half-order reaction rates with respect to oxygen concentration were found for UO 2 at all conditions tested. At room temperature, spent fuel dissolution (reaction) rates were nearly independent of oxygen concentration. At 75 degree C, reaction orders of 0.35 and 0.73 were observed for spent fuel, and there was some indication that the reaction order with respect to oxygen concentration might be dependent on pH and/or carbonate concentration as well as on temperature

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. Uranium-plutonium fuel for fast reactors

    International Nuclear Information System (INIS)

    Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

    1996-01-01

    Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

  2. 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.

  3. 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

  4. 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.

  5. 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…

  6. 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.)

  7. Water-Free Proton-Conducting Membranes for Fuel Cells

    Science.gov (United States)

    Narayanan, Sekharipuram; Yen, Shiao-Pin

    2007-01-01

    Poly-4-vinylpyridinebisulfate (P4VPBS) is a polymeric salt that has shown promise as a water-free proton-conducting material (solid electrolyte) suitable for use in membrane/electrode assemblies in fuel cells. Heretofore, proton-conducting membranes in fuel cells have been made from perfluorinated ionomers that cannot conduct protons in the absence of water and, consequently, cannot function at temperatures >100 C. In addition, the stability of perfluorinated ionomers at temperatures >100 C is questionable. However, the performances of fuel cells of the power systems of which they are parts could be improved if operating temperatures could be raised above 140 C. What is needed to make this possible is a solid-electrolyte material, such as P4VPBS, that can be cast into membranes and that both retains proton conductivity and remains stable in the desired higher operating temperature range. A family of solid-electrolyte materials different from P4VPBS was described in Anhydrous Proton-Conducting Membranes for Fuel Cells (NPO-30493), NASA Tech Briefs, Vol. 29, No. 8 (August 2005), page 48. Those materials notably include polymeric quaternized amine salts. If molecules of such a polymeric salt could be endowed with flexible chain structures, it would be possible to overcome the deficiencies of simple organic amine salts that must melt before being able to conduct protons. However, no polymeric quaternized amine salts have yet shown to be useful in this respect. The present solid electrolyte is made by quaternizing the linear polymer poly- 4-vinylpyridine (P4VP) to obtain P4VPBS. It is important to start with P4VP having a molecular weight of 160,000 daltons because P4VPBS made from lower-molecular-weight P4VP yields brittle membranes. In an experimental synthesis, P4VP was dissolved in methanol and then reacted with an excess of sulfuric acid to precipitate P4VPBS. The precipitate was recovered, washed several times with methanol to remove traces of acid, and dried to a

  8. 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

  9. Development of oxygen scavenger additives for jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beaver, B.D.; Demunshi, R.; Sharief, V.; Tian, D.; Teng, Y. [Duquesne Univ., Pittsburgh, PA (United States)

    1995-05-01

    Our current research program is in response to the US Air Force`s FY93 New Initiative entitled {open_quotes}Advanced Fuel Composition and Use.{close_quotes} The critical goal of this initiative is to develop aircraft fuels which can operate at supercritical conditions. This is a vital objective since future aircraft designs will transfer much higher heat loads into the fuel as compared with current heat loads. In this paper it is argued that the thermal stability of most jet fuels would be dramatically improved by the efficient in flight-removal of a fuel`s dissolved oxygen. It is proposed herein to stabilize the bulk fuel by the addition of an additive which will be judiciously designed and programmed to react with oxygen and produce an innocuous product. It is envisioned that a thermally activated reaction will occur, between the oxygen scavenging additive and dissolved oxygen, in a controlled and directed manner. Consequently formation of insoluble thermal degradation products will be limited. It is believed that successful completion of this project will result in the development of a new type of jet fuel additive which will enable current conventional jet fuels to obtain sufficient thermal stability to function in significantly higher temperature regimes. In addition, it is postulated that the successful development of thermally activated oxygen scavengers will also provide the sub-critical thermal stability necessary for future development of endothermic fuels.

  10. 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

  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. 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

  13. 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.

  14. 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

  15. An overview on dry reprocessing of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Ouyang Yinggen

    2002-01-01

    Although spent nuclear fuels have been reprocessed successfully for many years by the well-know Purex process based on solvent extraction, other reprocessing method which do not depend upon the use of organic solvents and aqueous media appear to have important potential advantage. There are two main non-aqueous methods for the reprocessing of spent fuel: fluoride-volatility process and pyro-electrochemical process. The presence of a poser in the process is that PuF 6 is obviously thermodynamically stable only in the presence of a large excess of fluorine. Pyro-electrochemical process is suited to processing metallic, oxide and carbide fuels. First, the fuel is dissolved in fresh salts, then, electrodes are introduced into the bath, U and Pu are deposited on the cathode, third, separation and refinement U and Pu are deposited on the cathode. There is a couple of contradictions in the process that are not in harmonious proportion in the fields on the nuclear fuel is dissolved the ability in the molten salt and corrosiveness of the molten salt for equipment used in the process

  16. Critical experiments with mixed oxide fuel

    International Nuclear Information System (INIS)

    Harris, D.R.

    1997-01-01

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

  17. 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

  18. Analytic solution of pseudocolloid migration in fractured rock

    International Nuclear Information System (INIS)

    Hwang, Y.; Pigford, T.H.; Lee, W.W.L.; Chambre, P.L.

    1989-06-01

    A form of colloid migration that can enhance or retard the migration of a dissolved contaminant in ground water is the sorption of the contaminant on the moving colloidal particulate to form pseudocolloids. In this paper we develop analytical solutions for the interactive migration of radioactive species dissolved in ground water and sorbed as pseudocolloids. The solute and pseudocolloids are assumed to undergo advection and dispersion in a one-dimensional flow field in planar fractures in porous rock. Interaction between pseudocolloid and dissolved species is described by equilibrium sorption. Sorbed species on the pseudocolloids undergo radioactive decay, and pseudocolloids can sorb on fracture surfaces and sediments. Filtration is neglected. The solute can decay and sorb on pseudocolloids, on the fracture surfaces, and on sediments and can diffuse into the porous rock matrix. 1 fig

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

    Ollila, K.

    2011-05-01

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

  2. 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.

  3. Removal of sulfamic acid from plutonium sulfamate--sulfamic acid solution

    International Nuclear Information System (INIS)

    Gray, L.W.

    1978-10-01

    Plutonium metal can be readily dissolved in aqueous solutions of sulfamic acid. When the plutonium sulfamate--sulfamic acid solutions are added to normal purex process streams, the sulfamate ion is oxidized by addition of sodium nitrite. This generates sodium sulfate which must be stored as radioactive waste. When recovery of ingrown 241 Am or storage of the dissolved plutonium must be considered, the sulfamate ion poses major and undesirable precipitation problems in the process streams. The present studies show that 40 to 80% of the sulfamate present in the dissolver solutions can be removed by precipitation as sulfamic acid by the addition of concentrated nitric acid. Addition of 64% nitric acid allows precipitation of 40 to 50% of the sulfamate; addition of 72% nitric acid allows precipitation of 50 to 60% of the sulfamate. If the solutions are chilled, additional sulfamic acid will precipitate. If the solutions are chilled to -10 0 C, about 70 to 80% of the orginal sulfamic acid in the dissolver will precipitate. A single, low-volume wash of the sulfamic acid crystals with concentrated nitric acid will decontaminate the crystals to a plutonium content of 5 dis/(min-gram)

  4. 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)

  5. 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

  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. 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

  8. 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...

  9. 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.

  10. Activity coefficients of solutes in binary solvents

    International Nuclear Information System (INIS)

    Gokcen, N.A.

    1982-01-01

    The activity coefficients in dilute ternary systems are discussed in detail by using the Margules equations. Analyses of some relevant data at high temperatures show that the sparingly dissolved solutes in binary solvents follow complex behavior even when the binary solvents are very nearly ideal. It is shown that the activity data on the solute or the binary system cannot permit computation of the remaining activities except for the regular solutions. It is also shown that a fourth-order equation is usually adequate in expressing the activity coefficient of a solute in binary solvents at high temperatures. When the activity data for a binary solvent are difficult to obtain in a certain range of composition, the activity data for a sparingly dissolved solute can be used to supplement determination of the binary activities

  11. Storage and management of fuel from fast breeder test reactor and KAlpakkam MINI

    International Nuclear Information System (INIS)

    Sodhi, B.S.; Rao, M.S.; Natarajan, R.

    1999-01-01

    Two Research Reactors, FBTR (Fast Breeder Test Reactor) and KAMINI (KAlpakkam MINI) are in operation at Kalpakkam, India. FBTR is a 40 MWt reactor. It is the first reactor to use mixed carbide (70% PuC-30% UC) as driver fuel. Special precautions are needed to fabricate pellets in glove boxes under inert atmosphere to take into account the possibility of criticality, radiation, pyrophoricity and toxicity of PuC. FBTR has been operating with small core up to 12 MWt power. The initial limit was 250 W/cm, linear heat rating and 25,000 MWd/t peak burnup. This limit was increased to 320 W/cm and 50,000 MWd/t respectively after rigorous analysis. At present the core has reached 40,000 MWd/t without any pin failure. After 25,000 MWd/t burnup one fuel subassembly (SA) was removed and PEE was carried out. The results were as expected by the analysis. In FBTR, fuel is stored in a container filled with argon and the container is cooled by forced circulation of air (during storage). Closing the fuel cycle is important for the breeder programme. Therefore, efforts have been made to set up a reprocessing plant. It uses the well proven purex process. The irradiated fuel is sheared in a single pin chopper and dissolved in an electrochemical dissolver. The resulting solution after adjusting the valency of Pu to IVth state is processed in the solvent extraction plant using 30% Tri-n-Butyl phosphate/n-dodecane as solvent. KAMINI is 30 kWt neutron source reactor which uses light water as moderator and coolant and has as a fuel U-233 aluminium alloy. Uranium-233 has been indigenously recovered from thorium irradiated in CIRUS reactor at Trombay. KAMINI was made critical on October 1996. It is housed in a vault below one of the hot cells in the Radiometallurgy laboratories of IGCAR. This reactor is planned to be used for neutron radiography of fuel elements and neutron activation analysis. It is available for use by research institutions and universities also. This paper describes the

  12. 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

  13. 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

  14. 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.

  15. 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)

  16. 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.

  17. Electrochemical methods to study hydrogen production during interaction of copper with deoxygenated aqueous solution

    International Nuclear Information System (INIS)

    Lilja, Christina; Betova, Iva; Bojinov, Martin

    2016-01-01

    In some countries, spent nuclear fuel is planned to be encapsulated in canisters with a copper shell for corrosion protection, for further disposal in geologic repositories. The possibilities for corrosion after oxygen depletion must be evaluated, even if copper is considered to be immune in oxygen-free water. To follow the interaction of copper with deoxygenated aqueous solution, open-circuit potentiometric and electrochemical impedance measurements have been coupled to in-situ detection of cupric ion, dissolved molecular hydrogen and oxygen concentrations using electrochemical sensors. A kinetic model that considers the production of hydrogen as a catalytic process, the rate of which is proportional to the surface coverage of an intermediate species formed during interaction between copper and the solution is used to interpret the results. Kinetic parameters are estimated by a simultaneous fit of the experimental impedance spectra, the open circuit potential and cupric ion concentration as depending on temperature (22–70 °C) and exposure time (up to 720 h) to the model equations. Using the obtained values and a balance equation of hydrogen production on copper and its diffusion out of the cell through its walls, the kinetic parameters of this process are estimated by fitting dissolved molecular hydrogen concentration vs. time data at the three temperatures.

  18. 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)

  19. Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

    International Nuclear Information System (INIS)

    Green, L.W.; Elliot, N.L.; Longhurst, T.H.

    1983-01-01

    A microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O 2 fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO 3 solution. The bead sorbed approximately 1.7 μg of U and 4.8μg of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO 3 solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < 0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. 6 figures, 2 tables

  20. Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Green, L.W.; Elliot, N.L.; Longhurst, T.H

    1983-07-01

    A convenient and sensitive microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O{sub 2} fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO{sub 3} solution. The bead sorbed approximately 1.7 {mu}g of U and 4.8 {mu}g of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO{sub 3} solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < -0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. (author)

  1. Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

    International Nuclear Information System (INIS)

    Fleming Rubio, Peter Alex

    2010-01-01

    The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

  2. Removal of radium from aqueous sulphate solutions

    International Nuclear Information System (INIS)

    Weir, D.R.; Masters, J.T.; Neven, M.

    1983-01-01

    Radium is often present in ores and an aqueous solution associated with the ore may consequently contain dissolved radium. It is frequently necessary to remove radium from such solutions to reduce the total radium content to a prescribed low level before the solution can be returned to the environment. The present invention is based on the discovery that the total radium content can be reduced to a satisfactory level within a reasonable time by adding a soluble barium salt to a radium-containing sulphate solution which also contains dissolved magnesium at a pH not greater than about 0 to precipitate radium as barium radium sulphate, raising the pH to at least 11 to precipitate an insoluble magnesium compound which collects the barium radium sulphate precipitate, and separating substantially all of the precipitates from the solution

  3. 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.

  4. 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

  5. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  6. 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....

  7. 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)

  8. 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.)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

    International Nuclear Information System (INIS)

    Heremanns, R.H.; Vandersteene, J.J.

    1983-01-01

    The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

  11. Improved measurement of aluminum in irradiated fuel reprocessed at the Savannah River Site

    International Nuclear Information System (INIS)

    Maxwell, S.L. III.

    1991-01-01

    At the Savannah River Site (SRS), irradiated fuel from research reactor operators or their contract fuel service companies is reprocessed in the H-Canyon Separations Facility. Final processing costs are based on analytical measurements of the amount of total metal dissolved. Shipper estimates for uranium and uranium-235 and measured values at SRS have historically agreed very well. There have occasionally been significant differences between shipper estimates for aluminum and the aluminum content determined at SRS. To minimize analytical error that might contribute to poor shipper-receiver agreement for the reprocessing of off-site fuel, a new analytical method to measure aluminum was developed by SRS Analytical Laboratories at the Central Laboratory Facilities. An EDTA (ethylenediaminetetraacetic acid) titration method, subject to dissolver matrix interferences, was previously used at SRS to measure aluminum in H-Canyon dissolver during the reprocessing of offsite fuel. The new method combines rapid ion exchange technology with direct current argon plasma spectrometry to enhance the reliability of aluminum measurements for off-site fuel. The technique rapidly removes spectral interferences such as uranium and significantly lowers gamma levels due to fission products. Aluminium is separated quantitatively by using an anion exchange technique that employs oxalate complexing, small particle size resin and rapid flow rates. The new method, which has eliminated matrix interference problems with these analyses and improved the quality of aluminum measurements, has improved the overall agreement between shipper-receiver values for offsite fuel processed SRS

  12. 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

  13. 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.

  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. 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)

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.)

  1. Predicting nitrogen and acidity effects on long-term dynamics of dissolved organic matter

    International Nuclear Information System (INIS)

    Rowe, E.C.; Tipping, E.; Posch, M.; Oulehle, F.; Cooper, D.M.; Jones, T.G.; Burden, A.; Hall, J.; Evans, C.D.

    2014-01-01

    Increases in dissolved organic carbon (DOC) fluxes may relate to changes in sulphur and nitrogen pollution. We integrated existing models of vegetation growth and soil organic matter turnover, acid–base dynamics, and organic matter mobility, to form the ‘MADOC’ model. After calibrating parameters governing interactions between pH and DOC dissolution using control treatments on two field experiments, MADOC reproduced responses of pH and DOC to additions of acidifying and alkalising solutions. Long-term trends in a range of acid waters were also reproduced. The model suggests that the sustained nature of observed DOC increases can best be explained by a continuously replenishing potentially-dissolved carbon pool, rather than dissolution of a large accumulated store. The simulations informed the development of hypotheses that: DOC increase is related to plant productivity increase as well as to pH change; DOC increases due to nitrogen pollution will become evident, and be sustained, after soil pH has stabilised. -- Highlights: • A model of dissolved organic carbon (DOC) was developed by integrating simple models • MADOC simulates effects of sulphur and nitrogen deposition and interactions with pH. • Responses of DOC and pH to experimental acidification and alkalisation were reproduced. • The persistence of DOC increases will depend on continued supply of potential DOC. • DOC fluxes are likely determined by plant productivity as well as soil solution pH. -- Effects of changes in sulphur and nitrogen pollution on dissolved organic carbon fluxes are predicted by simulating soil organic matter cycling, the release of potentially-dissolved carbon, and interactions with soil pH

  2. 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)

  3. 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)

  4. 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

  5. Development of remote maintenance technology for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Kawahara, Akira; Saito, Masayuki; Kawamura, Hironobu; Yamade, Atsushi; Sugiyama, Sen; Sugiyama, Sakae.

    1986-01-01

    In the plants for reprocessing spent nuclear fuel containing fission products, due to the facts that the facilities are in high radiations fields, and the surfaces of equipments are contaminated with radioactive substances, the troubles of process equipments are directly connected to the remarkable drop of the rate of operation of the facilities. Therefore, the development of various remote maintenance techniques has been carried out so far, but this time, Hitachi Ltd. got a chance to take part in the repair of spent fuel dissolving tanks in the Tokai Reprocessing Plant of Power Reactor and Nuclear Fuel Development Corp. and the development of several kinds of remote checkup equipment related to the repair work. Especially in the repair of the dissolving tanks, a radiation-withstanding checkup and repair apparatus which has high remote operability taking the conditions of radioactive environment and the restriction of the repaired objects in consideration was required, and a dissolving tank repairing robot composed of six kinds has been developed. The key points of the development were the selective use of high radiation-withstanding parts and materials, small size structure and the realization of full remote operability. The full remote maintenance apparatus of this kind is unique in the world, and applicable to wide fields. (Kako, I.)

  6. Do Regional Aerosols Contribute to the Riverine Export of Dissolved Black Carbon?

    Science.gov (United States)

    Jones, M. W.; Quine, T. A.; de Rezende, C. E.; Dittmar, T.; Johnson, B.; Manecki, M.; Marques, J. S. J.; de Aragão, L. E. O. C.

    2017-11-01

    The fate of black carbon (BC), a stable form of thermally altered organic carbon produced during biomass and fuel combustion, remains an area of uncertainty in the global carbon cycle. The transfer of photosynthetically derived BC into extremely long-term oceanic storage is of particular significance and rivers are the key linkage between terrestrial sources and oceanic stores. Significant fluvial fluxes of dissolved BC to oceans result from the slow release of BC from degrading charcoal stocks; however, these fluvial fluxes may also include undetermined contributions of aerosol BC, produced by biomass and fossil fuel combustion, which are deposited in river catchments following atmospheric transport. By investigation of the Paraíba do Sul River catchment in Southeast Brazil we show that aerosol deposits can be substantial contributors to fluvial fluxes of BC. We derived spatial distributions of BC stocks within the catchment associated with soil charcoal and with aerosol from both open biomass burning and fuel combustion. We then modeled the fluvial concentrations of dissolved BC (DBC) in scenarios with varying rates of export from each stock. We analyzed the ability of each scenario to reproduce the variability in DBC concentrations measured in four data sets of river water samples collected between 2010 and 2014 and found that the best performing scenarios included a 5-18% (135-486 Mg DBC year-1) aerosol contribution. Our results suggest that aerosol deposits of BC in river catchments have a shorter residence time in catchments than charcoal BC and, therefore, contribute disproportionately (with respect to stock magnitude) toward fluvial fluxes of BC.

  7. 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

  8. 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

  9. The Dissolution of Uranium Oxides in HB-Line Phase 1 Dissolvers

    International Nuclear Information System (INIS)

    Gray, J.H.

    2003-01-01

    A series of characterization and dissolution studies has been performed to define flowsheet conditions for the dissolution of uranium oxide materials in dissolvers. The samples selected for analysis were uranium oxide materials. The selection of these uranium oxide materials for characterization and dissolution studies was based on high enriched uranium content and trace levels of plutonium. Test results from the characterization study identified ferric oxide (Fe2O3) and iron/chromium/nickel (Fe/Cr/Ni) particles as impurities along with the tri-uranium oxide (U3O8) and uranium trioxide (UO3). The weight percent uranium in this material was found to vary depending on the impurity content. The trace impurity plutonium appears to be associated with the Fe/Cr/Ni particles. A small amount of absorbed moisture and waters of hydration is present. Most of the uranium oxides easily dissolved in low-molar nitric acid solutions without fluoride within one to two hours at solution temperature s between 60-80 degrees C. A small amount of residue remained following this dissolution step. To assure complete dissolution of uranium from these oxide materials, an additional dissolution step at 90 degrees C to boiling for at least one to two hours has been suggested. Only trace amounts of iron associated with Fe2O3 and Fe/Cr/Ni particles will dissolve during the dissolution steps. Neither hydrogen nor heat will be generated during the dissolution of these uranium oxide materials in nitric acid solutions. Some brown nitrogen dioxide (NO2) fumes will be generated during the dissolution of U3O8

  10. 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.

  11. 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.)

  12. Small-scale irradiated fuel electrorefining

    International Nuclear Information System (INIS)

    Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

    1993-01-01

    In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

  13. Pyroelectrochemical process for reprocessing irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1982-01-01

    A pyroelectrochemical process for reprocessing irradiated fast reactor mixed oxide or carbide fuels is described. The fuel is dissolved in a bath of molten alkali metal sulfates. The Pu(SO 4 ) 2 formed in the bath is thermally decomposed, leaving crystalline PuO 2 on the bottom of the reaction vessel. Electrodes are then introduced into the bath, and UO 2 is deposited on the cathode. Alternatively, both UO 2 and PuO 2 may be electrodeposited. The molten salts, after decontamination by precipitating the fission products dissolved in the bath by introducing basic agents such as oxides, carbonates, or hydroxides, may be recycled. Since it is not possible to remove cesium from the molten salt bath, periodic disposal and partial renewal with fresh salts is necessary. The melted salts that contain the fission products are conditioned for disposal by embedding them in a metallic matrix

  14. Recovery of uranium from (U,Gd)O{sub 2} nuclear fuel scrap using dissolution and precipitation in carbonate media

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang-Wook, E-mail: nkwkim@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok daero, Yuseong, Daejeon 305-353 (Korea, Republic of); KEPCO NF 1047 Daedeok daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Hyun, Jun-Taek; Lee, Eil-Hee; Park, Geun-Il; Lee, Kune-Woo [Korea Atomic Energy Research Institute, 1045 Daedeok daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Yoo, Myung-June [KEPCO NF 1047 Daedeok daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Song, Kee-Chan; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, 1045 Daedeok daero, Yuseong, Daejeon 305-353 (Korea, Republic of)

    2011-11-15

    Highlights: > A treatment of (U,Gd)O{sub 2} scrap with a dissolution in carbonate solution with H{sub 2}O{sub 2}. > Partial dissolution of Gd together with uranium in carbonate solution. > Solubilities of Gd in solutions with and without carbonate at several pHs. > Purification of Gd-contaminated UO{sub 4} by dissolution and precipitation of UO{sub 4}. - Abstract: This work studied a process to recover uranium from contaminated (U,Gd)O{sub 2} scraps generated from nuclear fuel fabrication processes by using the dissolution of (U,Gd)O{sub 2} scraps in a carbonate with H{sub 2}O{sub 2} and the precipitation of the dissolved uranium as UO{sub 4}. The dissolution characteristics of uranium, Gd, and impurity metal oxides were tested, and the behaviors of UO{sub 4} precipitation and Gd solubility were evaluated with changes of the pH of the solution. A little Gd was entrained in the UO{sub 4} precipitate to contaminate the uranium precipitate. Below a pH of 3, the uranium dissolved in the form of uranyl peroxo-carbonato complex ions in the carbonate solution was precipitated as UO{sub 4} with a high precipitation yield, and the Gd had a very high solubility. Using these characteristics, the Gd-contaminated UO{sub 4} could be purified using dissolution in a 1-M HNO{sub 3} solution with heating and re-precipitation upon addition of H{sub 2}O{sub 2} to the solution. Finally, an environmentally friendly and economical process to recover pure uranium from contaminated (U,Gd)O{sub 2} scraps was suggested.

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. Application of fuel cell for pyrite and heavy metal containing mining waste

    Science.gov (United States)

    Keum, H.; Ju, W. J.; Jho, E. H.; Nam, K.

    2015-12-01

    Once pyrite and heavy metal containing mining waste reacts with water and air it produces acid mine drainage (AMD) and leads to the other environmental problems such as contamination of surrounding soils. Pyrite is the major source of AMD and it can be controlled using a biological-electrochemical dissolution method. By enhancing the dissolution of pyrite using fuel cell technology, not only mining waste be beneficially utilized but also be treated at the same time by. As pyrite-containing mining waste is oxidized in the anode of the fuel cell, electrons and protons are generated, and electrons moves through an external load to cathode reducing oxygen to water while protons migrate to cathode through a proton exchange membrane. Iron-oxidizing bacteria such as Acidithiobacillus ferrooxidans, which can utilize Fe as an electron donor promotes pyrite dissolution and hence enhances electrochemical dissolution of pyrite from mining waste. In this study mining waste from a zinc mine in Korea containing 17 wt% pyrite and 9% As was utilized as a fuel for the fuel cell inoculated with A. ferrooxidans. Electrochemically dissolved As content and chemically dissolved As content was compared. With the initial pH of 3.5 at 23℃, the dissolved As concentration increased (from 4.0 to 13 mg/L after 20 d) in the fuel cell, while it kept decreased in the chemical reactor (from 12 to 0.43 mg/L after 20 d). The fuel cell produced 0.09 V of open circuit voltage with the maximum power density of 0.84 mW/m2. Dissolution of As from mining waste was enhanced through electrochemical reaction. Application of fuel cell technology is a novel treatment method for pyrite and heavy metals containing mining waste, and this method is beneficial for mining environment as well as local community of mining areas.

  20. 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

  1. Equipment specifications for an electrochemical fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hemphill, Kevin P.

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  2. 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.

  3. Novel materials for fuel cells operating on liquid fuels

    Directory of Open Access Journals (Sweden)

    César A. C. Sequeira

    2017-05-01

    Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

  4. Electrochemical reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    Brambilla, G.; Sartorelli, A.

    1980-01-01

    A method is described for the reprocessing of irradiated nuclear fuel which is particularly suitable for use with fuel from fast reactors and has the advantage of being a dry process in which there is no danger of radiation damage to a solvent medium as in a wet process. It comprises the steps of dissolving the fuel in a salt melt under such conditions that uranium and plutonium therein are converted to sulphate form. The plutonium sulphate may then be thermally decomposed to PuO 2 and removed. The salt melt is then subjected to electrolysis conditions to achieve cathodic deposition of UO 2 (and possibly PuO 2 ). The salt melt can then be recycled or conditioned for final disposal. (author)

  5. 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.

  6. 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

  7. Removal of 137-Cs from Dissolved Hanford Tank Saltcake by Treatment with IE-911

    International Nuclear Information System (INIS)

    Rapko, Brian M.; Sinkov, Sergei I.; Levitskaia, Tatiana G.

    2003-01-01

    The U.S. Department of Energy's Richland Operations Office plans to accelerate the cleanup of the Hanford Site. Testing new technology for the accelerated cleanup will require dissolved saltcake from single-shell tanks. However, the 137Cs will need to be removed from the saltcake to alleviate radiation hazards. A saltcake composite constructed from archived samples from Hanford Site single-shell tanks 241-S-101, 241-S-109, 241-S-110, 241-S-111, 241-U-106, and 241-U-109 was dissolved in water, adjusted to 5 M Na, and transferred from the 222-S Laboratory to the Radiochemical Processing Laboratory (RPL). At the RPL, the approximately 5.5 liters of solution was passed through a 0.2-micron polyethersulfone filter, collected, and homogenized. The filtered solution then was passed through an ion exchange column containing approximately 150 mL IONSIV(reg s ign) IE-911, an engineered form of crystalline silicotitanate available from UOP, at approximately 200 mL/hour in a continuous operation until all of the feed solution had been run through the column. An analysis of the 137Cs concentrations in the initial feed solution and combined column effluent indicates that > 99.999 percent of the Cs in the feed solution was removed by this operation

  8. 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

  9. Behavior of silicon in nitric media. Application to uranium silicides fuels reprocessing; Comportement du silicium en milieu nitrique. Application au retraitement des combustibles siliciures d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Cheroux, L

    2001-07-01

    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)

  10. 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 ...

  11. 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

  12. Solutions of group IV elements in liquid lithium

    International Nuclear Information System (INIS)

    Dadd, A.T.; Hubberstey, P.; Roberts, P.G.

    1982-01-01

    The solubilities of tin (0.00 = 22 Sn 5 . A simple thermochemical cycle is used to demonstrate that, whereas carbon dissolves endothermically in both liquid lithium and liquid sodium, the heavier Group IV elements dissolve exothermically. A similar cycle is used to derive solvation enthalpies (for the neutral gaseous species) for all Group IV elements in the two solvents. The trend in solvation enthalpy: C > Si > Ge > Sn > Pb is indicative of a diminishing affinity of solvent for solute and is attributed to the increasing metallic character of the solute as the Group is descended. (author)

  13. 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

  14. Dissolution of nuclear fuel samples for analytical purposes. I

    International Nuclear Information System (INIS)

    Krtil, J.

    1983-01-01

    Main attention is devoted to procedures for dissolving fuels based on uranium metal and its alloys, uranium oxides and carbides, plutonium metal, plutonium dioxide, plutonium carbides, mixed PuC-UC carbides and mixed oxides (PuU)O 2 . Data from the literature and experience gained with the dissolution of nuclear fuel samples at the Central Control Laboratory of the Nuclear Research Institute at Rez are given. (B.S.)

  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. 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

  17. Fast dissolving strips: A novel approach for the delivery of verapamil

    Directory of Open Access Journals (Sweden)

    S Kunte

    2010-01-01

    Full Text Available Objective: Fast dissolving drug delivery system offers a solution for those patients having difficulty in swallowing tablets/capsules etc. Verapamil is a calcium channel blocker used as an antianginal, antiarrhythmic, and antihypertensive agent with extensive first pass metabolism which results in less bioavailability. This work investigated the possibility of developing verapamil fast dissolving strips allowing fast, reproducible drug dissolution in the oral cavity; thus bypassing first pass metabolism. Materials and methods: The fast dissolving strips were prepared by solvent casting technique with the help of HPMC E6 and maltodextrin. The strips were evaluated for drug content uniformity, film thickness, folding endurance, in vitro disintegration time, in vitro dissolution studies, surface pH study, and palatability study. Results: Official criteria for evaluation parameters were fulfilled by all formulations. Disintegration time showed by formulations was found to be in range of 20.4-28.6 sec. Based on the evaluation parameters, the formulation containing 2% HPMC E6 and 3.5% maltodextrin showed optimum performance against other formulations. Conclusion: It was concluded that the fast dissolving strips of verapamil can be made by solvent casting technique with enhanced dissolution rate, taste masking, and hence better patient compliance and effective therapy

  18. 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

  19. History and current status of nuclear fuel reprocessing technology

    International Nuclear Information System (INIS)

    Funasaka, Hideyuki; Nagai, Toshihisa; Washiya, Tadahiro

    2008-01-01

    History and present state of fast breeder reactor was reviewed in series. As a history and current status of nuclear fuel reprocessing technology, this ninth lecture presented the progress of the FBR fuel reprocessing technology and advanced reprocessing processes. FBR fuel reprocessing technology had been developed to construct the reprocessing equipment test facilities (RETF) based on PUREX process technologies. With economics, reduction of environmental burdens and proliferation resistance taken into consideration, advanced aqueous method for nuclear fuel cycle activities has been promoted as the government's basic policy. Innovative technologies on mechanical disassembly, continuous rotary dissolver, crystallizer, solvent extraction and actinides recovery have been mainly studied. (T. Tanaka)

  20. Post-Irradiation Examinations for Resolving Fuel Issues in Long Term Storage

    International Nuclear Information System (INIS)

    Karlsson, Joakim K.H.; Alvarez Holston, Anna-Maria

    2014-01-01

    In many countries extended long term dry storage is the solution for storage of spent nuclear fuel for the foreseeable future. The expected storage times have increased over the last years and today storage times of up to 300 years is anticipated. With such long storage times, requirements on transportability and retrievability of the fuel have become more important. Hitherto most investigations on fuel behaviour during dry storage have been focused on cladding creep and the impact of hydrogen and hydrides in the cladding. Creep data gives input to creep models and creep to rupture data helps to set criteria for maximum allowable internal rod pressure. Hydrides lower the ductility of the cladding and this is more pronounced with radially oriented hydrides. As the temperature decreases over time in a dry storage cask dissolved hydrogen will precipitate forming hydrides in addition to hydrides already present. Assuming there is sufficient hoop stress in the cladding, the new hydrides would be radially oriented. Together with lost ductility Delayed Hydride Cracking (DHC) could be a potential mechanism for rod failure over tens of years of dry storage as the temperature drops from about 350 deg. C to 150 deg. C. Hydride embrittlement and the DHC mechanism have been studied in the first Studsvik Cladding Integrity Project (SCIP), although the focus in this program has mainly been on higher temperatures relevant for operating conditions rather than on dry storage conditions. In addition to the mechanisms mentioned there are other failure mechanisms that could potentially threaten the cladding fuel integrity and retrievability. In case there is residual water or moisture available in the cask, or even in the fuel due to existing fuel failures, radiolysis gives free hydrogen and oxygen. In failed fuel this may cause fuel oxidation and swelling affecting fuel integrity. The hydrogen gas pressure will not threaten the cask but be available for cladding uptake. Furthermore

  1. 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.

  2. A flow injection analyser conductometric coupled system for the field analysis of free dissolved CO{sub 2} and total dissolved inorganic carbon in natural waters

    Energy Technology Data Exchange (ETDEWEB)

    Martinotti, Valter; Balordi, Marcella; Ciceri, Giovanni [RSE SpA - Environment and Sustainable Development Department, Milan (Italy)

    2012-05-15

    A flow injection analyser coupled with a gas diffusion membrane and a conductometric microdetector was adapted for the field analysis of natural concentrations of free dissolved CO{sub 2} and dissolved inorganic carbon in natural waters and used in a number of field campaigns for marine water monitoring. The dissolved gaseous CO{sub 2} presents naturally, or that generated by acidification of the sample, is separated by diffusion using a hydrophobic semipermeable gas porous membrane, and the permeating gas is incorporated into a stream of deionised water and measured by means of an electrical conductometric microdetector. In order to make the system suitable and easy to use for in-field measurements aboard oceanographic ships, the single components of the analyser were compacted into a robust and easy to use system. The calibration of the system is carried out by using standard solutions of potassium bicarbonate at two concentration ranges. Calibration and sample measurements are carried out inside a temperature-constant chamber at 25 C and in an inert atmosphere (N{sub 2}). The detection and quantification limits of the method, evaluated as 3 and 10 times the standard deviation of a series of measurements of the matrix solution were 2.9 and 9.6 {mu}mol/kg of CO{sub 2}, respectively. Data quality for dissolved inorganic carbon was checked with replicate measurements of a certified reference material (A. Dickson, Scripps Institution of Oceanography, University of California, San Diego), both accuracy and repeatability were -3.3% and 10%, respectively. Optimization, performance qualification of the system and its application in various natural water samples are reported and discussed. In the future, the calibration step will be operated automatically in order to improve the analytical performance and the applicability will be increased in the course of experimental surveys carried out both in marine and freshwater ecosystems. Considering the present stage of

  3. Dissolved stable noble gas measurements from primary water of Paks NPP

    International Nuclear Information System (INIS)

    Palcsu, L.; Molnar, M.; Szanto, Zs.; Svingor, E.; Futo, I.; Pinter, T.

    2001-01-01

    A sampling and measuring method of noble gases from the primary water circuit of a VVER type NPP was developed to provide relevant information about the kilter of heating rods and detailed additional information about some working parameters. The helium concentrations and 3 He/ 4 He ratios was used to estimate the content of tritium and alpha emitting isotopes of the primary water. By argon content measurements the air penetration and the required hydrazine amount for the oxygen absorption could be estimated with high accuracy. Continuous monitoring of the concentration and isotope ratios of Xe and Kr in the dissolved gas is proved to be a good tool for high sensitivity detection of small leakage of fuel elements. In case of block-3 xenon surplus was detected. The results indicate possible leakage of fuel rods.(author)

  4. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  5. A comparison of sodium borohydride as a fuel for proton exchange membrane fuel cells and for direct borohydride fuel cells

    Science.gov (United States)

    Wee, Jung-Ho

    Two types of fuel cell systems using NaBH 4 aqueous solution as a fuel are possible: the hydrogen/air proton exchange membrane fuel cell (PEMFC) which uses onsite H 2 generated via the NaBH 4 hydrolysis reaction (B-PEMFC) at the anode and the direct borohydride fuel cell (DBFC) system which directly uses NaBH 4 aqueous solution at the anode and air at the cathode. Recently, research on these two types of fuel cells has begun to attract interest due to the various benefits of this liquid fuel for fuel cell systems for portable applications. It might therefore be relevant at this stage to evaluate the relative competitiveness of the two fuel cells. Considering their current technologies and the high price of NaBH 4, this paper evaluated and analyzed the factors influencing the relative favorability of each type of fuel cell. Their relative competitiveness was strongly dependent on the extent of the NaBH 4 crossover. When considering the crossover in DBFC systems, the total costs of the B-PEMFC system were the most competitive among the fuel cell systems. On the other hand, if the crossover problem were to be completely overcome, the total cost of the DBFC system generating six electrons (6e-DBFC) would be very similar to that of the B-PEMFC system. The DBFC system generating eight electrons (8e-DBFC) became even more competitive if the problem of crossover can be overcome. However, in this case, the volume of NaBH 4 aqueous solution consumed by the DBFC was larger than that consumed by the B-PEMFC.

  6. 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

  7. New measurement capabilities of mass spectrometry in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Perrin, R.E.

    1979-01-01

    Three recent developments, when combined, have the potential for greatly improving accountability measurements in the nuclear fuel cycle. The techniques are particularly valuable when measuring the contents of vessels which are difficult to calibrate by weight or volume. Input dissolver accountability measurements, inparticular, benefit from the application of these techniques. Los Alamos Scientific Laboratory has developed the capability for isotopic analysis of U and Pu samples at the nanogram level with an accuracy of 0.1 relative %. The Central Bureau for Nuclear Materials Measurement in Geel, Belgium has developed the capability of preparing mixed, solid metal U and Pu spikes with an accuracy of better than 0.1 relative %. Idaho Nuclear Energy Laboratory and C.K. Mathews at Bhabha Atomic Research have demonstrated a technique for determining the ratio of sample size to total solution measured which is independent of both the weight and the volume of the solution being measured. The advantages and limitations of these techniques are discussed. An analytical scheme which takes advantage of the special features of these techniques is proposed. 4 refs

  8. 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.

  9. 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.

  10. 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

  11. Determination of the thermodynamic state of concentrated hemoglobin solutions by means of small angle X-ray scattering

    International Nuclear Information System (INIS)

    Zinke, M.

    1979-01-01

    Exemplified by hemoglobin, the thermodynamic equilibrium properties of the dissolved macromolecular system could be determined solely from the small angle X-ray scattering of concentrated macromolecular solutions via the intermolecular structure of the dissolved macromolecules and their intermolecular potentials. From the scattering experiment on concentrated Hb solutions the concentration dependence of the following properties of the dissolved Hb system were determined: fluctuation, isothermic compressibility, internal energy, surface tension, and osmotic pressure. (author)

  12. 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)

  13. Programmatic Life Cycle Environmental Assessment for Smoke/Obscurants. Volume 1. Fog Oil, Diesel Fuels, and Polyethylene Glycol (PEG 200)

    Science.gov (United States)

    1983-07-01

    distances using TLV on these models requires conversion from concentration to dosage. The TWA (time weighted averae) for healthy adult humans exposed to oil...diseases. Chronic industrial exposures of oils and oil mists have been implicated in causing dermatosis (SGF No. 1) and dermatosis plus tumors of skin...mg/I No. 2 2-Day LCo Is Quahaug larvae fuel oil dissolved in water Exposure to 0.53 mg/l No. 10-Day LCo0 2 fuel oil dissolved in water Young adult

  14. Modification of an enzyme electrode by electrodeposition of hydroquinone for use as the anode of a glucose fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Kuwahara, Takashi; Yamazaki, Hiraku; Kondo, Mizuki [Department of Bioengineering, Faculty of Engineering, Nagaoka University of Technology, 1603-1, Kamitomioka-machi, Nagaoka 940-2188 (Japan); Shimomura, Masato, E-mail: smasato@vos.nagaokaut.ac.jp [Department of Bioengineering, Faculty of Engineering, Nagaoka University of Technology, 1603-1, Kamitomioka-machi, Nagaoka 940-2188 (Japan)

    2012-06-15

    An electrode having immobilized glucose oxidase (GOx) was modified with polyhydroquinone (PHQ), which was employed as an electron-transferring mediator, by a simple electrochemical method and used as the anode of a glucose fuel cell. The GOx-immobilized electrode was fabricated by attaching polyallylamine (PAAm) and then GOx covalently onto a gold electrode covered with a monolayer formed with 3-mercaptopropionic acid. Subsequently, the GOx-immobilized electrode (GOx/PAAm electrode) was modified with PHQ by electrodeposition of hydroquinone. The cyclic voltammogram of the modified electrode (PHQ/GOx/PAAm electrode) in a phosphate buffer solution (0.10 M, pH 7.0) showed redox peaks due to the electrodeposited PHQ, whereas no redox peaks were found for the GOx/PAAm electrode in the buffer solution containing p-benzoquinone (BQ). The onset potential of glucose oxidation with the PHQ/GOx/PAAm electrode became ca. 0.2 V more negative than that observed with the GOx/PAAm electrode in the presence of BQ. The glucose fuel cell equipped with the PHQ/GOx/PAAm electrode as an anode gave a 3 times larger power output than the cell with the GOx/PAAm electrode using dissolved quinone as the mediator.

  15. Surfactive stabilization of multi-walled carbon nanotube dispersions with dissolved humic substances

    Energy Technology Data Exchange (ETDEWEB)

    Chappell, Mark A. [Environmental Laboratory, Engineering Research and Development Center, US Army Corps of Engineers, 3909 Halls Ferry Road, Vicksburg, MS 39180 (United States)], E-mail: mark.a.chappell@usace.army.mil; George, Aaron J.; Dontsova, Katerina M.; Porter, Beth E. [SpecPro, Inc., 4815 Bradford Drive, Suite 201, Huntsville, AL 35805 (United States); Price, Cynthia L. [Environmental Laboratory, Engineering Research and Development Center, US Army Corps of Engineers, 3909 Halls Ferry Road, Vicksburg, MS 39180 (United States); Zhou Pingheng; Morikawa, Eizi [J. Bennett Johnston Sr. Center for Advanced Microstructures and Devices, Louisiana State University, 6980 Jefferson Highway, Baton Rouge, LA 70806 (United States); Kennedy, Alan J.; Steevens, Jeffery A. [Environmental Laboratory, Engineering Research and Development Center, US Army Corps of Engineers, 3909 Halls Ferry Road, Vicksburg, MS 39180 (United States)

    2009-04-15

    Soil humic substances (HS) stabilize carbon nanotube (CNT) dispersions, a mechanism we hypothesized arose from the surfactive nature of HS. Experiments dispersing multi-walled CNT in solutions of dissolved Aldrich humic acid (HA) or water-extractable Catlin soil HS demonstrated enhanced stability at 150 and 300 mg L{sup -1} added Aldrich HA and Catlin HS, respectively, corresponding with decreased CNT mean particle diameter (MPD) and polydispersivity (PD) of 250 nm and 0.3 for Aldrich HA and 450 nm and 0.35 for Catlin HS. Analogous trends in MPD and PD were observed with addition of the surfactants Brij 35, Triton X-405, and SDS, corresponding to surfactant sorption maximum. NEXAFS characterization showed that Aldrich HA contained highly surfactive domains while Catlin soil possessed a mostly carbohydrate-based structure. This work demonstrates that the chemical structure of humic materials in natural waters is directly linked to their surfactive ability to disperse CNT released into the environment. - Suspensions of multi-walled carbon nanotubes are stabilized by relatively low concentrations of dissolved humic substances in solution through surfactive mechanisms.

  16. Surfactive stabilization of multi-walled carbon nanotube dispersions with dissolved humic substances

    International Nuclear Information System (INIS)

    Chappell, Mark A.; George, Aaron J.; Dontsova, Katerina M.; Porter, Beth E.; Price, Cynthia L.; Zhou Pingheng; Morikawa, Eizi; Kennedy, Alan J.; Steevens, Jeffery A.

    2009-01-01

    Soil humic substances (HS) stabilize carbon nanotube (CNT) dispersions, a mechanism we hypothesized arose from the surfactive nature of HS. Experiments dispersing multi-walled CNT in solutions of dissolved Aldrich humic acid (HA) or water-extractable Catlin soil HS demonstrated enhanced stability at 150 and 300 mg L -1 added Aldrich HA and Catlin HS, respectively, corresponding with decreased CNT mean particle diameter (MPD) and polydispersivity (PD) of 250 nm and 0.3 for Aldrich HA and 450 nm and 0.35 for Catlin HS. Analogous trends in MPD and PD were observed with addition of the surfactants Brij 35, Triton X-405, and SDS, corresponding to surfactant sorption maximum. NEXAFS characterization showed that Aldrich HA contained highly surfactive domains while Catlin soil possessed a mostly carbohydrate-based structure. This work demonstrates that the chemical structure of humic materials in natural waters is directly linked to their surfactive ability to disperse CNT released into the environment. - Suspensions of multi-walled carbon nanotubes are stabilized by relatively low concentrations of dissolved humic substances in solution through surfactive mechanisms

  17. Further development of IDGS: Isotope dilution gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Li, T.K.; Parker, J.L.; Kuno, Y.; Sato, S.; Kamata, M.; Akiyama, T.

    1991-01-01

    The isotope dilution gamma-ray spectrometry (IDGS) technique for determining the plutonium concentration and isotopic composition of highly radioactive spent-fuel dissolver solutions has been further developed. Both the sample preparation and the analysis have been improved. The plutonium isotopic analysis is based on high-resolution, low-energy gamma-ray spectrometry. The plutonium concentration in the dissolver solutions then is calculated from the measured isotopic differences among the spike, the dissolver solution, and the spiked dissolver solution. Plutonium concentrations and isotopic compositions of dissolver solutions analyzed from this study agree well with those obtained by traditional isotope dilution mass spectrometry (IDMS) and are consistent with the first IDGS experimental result. With the current detector efficiency, sample size, and a 100-min count time, the estimated precision is ∼0.5% for 239 Pu and 240 Pu isotopic analyses and ∼1% for the plutonium concentration analysis. 5 refs., 2 figs., 7 tabs

  18. 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.)

  19. Dissolution test for homogeneity of mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

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

  20. Solutal convection induced by dissolution. Influence on erosion dynamics and interface shaping.

    Science.gov (United States)

    Berhanu, Michael; Philippi, Julien; Cohen, Caroline; Derr, Julien; Courrech du Pont, Sylvain

    2017-04-01

    Rock fractures invaded by a water flow, are often subjected to dissolution, which let grow and evolve the initial fracture network, by evacuating the eroded minerals under a solute form. In the case of fast kinetic of dissolution, local erosion rate is set by the advection of the solute. The erosion velocity decreases indeed with the solute concentration at the interface and vanishes when this concentration reaches the saturation value. Even in absence of an imposed or external flow, advection can drive the dissolution, when buoyancy effects due to gravity induce a solutal convection flow, which controls the erosive dynamics and modifies the shape of the dissolving interface. Here, we investigate using model experiments with fast dissolving materials and numerical simulations in simplified situations, solutal convection induced by dissolution. Results are interpreted regarding a linear stability analysis of the corresponding solutal Rayleigh-Benard instability. A dissolving surface is suspended above a water height, initially at rest. In a first step, solute flux is transported through a growing diffusion layer. Then after an onset time, once the layer exceeds critical width, convection flow starts under the form of falling plumes. A dynamic equilibrium results in average from births and deaths of intermittent plumes, setting the size of the solute concentration boundary layer at the interface and thus the erosion velocity. Solutal convection can also induce a pattern on the dissolving interface. We show experimentally with suspended and inclined blocks of salt and sugar, that in a linear stage, the first wavelength of the dissolution pattern corresponds to the wavelength of the convection instability. Then pattern evolves to more complex shapes due to non-linear interactions between the flow and the eroded interface. More generally, we inquire what are the conditions to observe a such solutal convection instability in geological situations and if the properties of

  1. The isolation of lutetium from gadolinium contained in Purex process solutions

    International Nuclear Information System (INIS)

    Bostick, D.T.; Vick, D.O.; May, M.P.; Walker, R.L.

    1992-09-01

    A chemical separation procedure has been devised to isolate Lu from Purex dissolver solutions containing the neutron poison, Gd. The isolation procedure involves the removal of U and >Pu from a dissolver solution using tributylphosphate solvent extraction. If required, solvent extraction using di-(2-ethylhexyl) phosphoric acid can be employed to further purify the sample be removing alkali and alkali earth elements. Finally, Lu is chromatographically separated from Gd and rare earth fission products on a Dowex 50W-X8 resin column using an alpha-hydroxyisobutyrate eluant. The success of the chemical separation procedure has been demonstrated in the quantitative recovery of as little as 1.4 ng Lu from solutions containing a 5000-fold excess of Gd. Additionally, Lu has been isolated from synthetic dissolver samples containing U, Ba, Cs, and Gd. Thermal emission MS data indicated that the Lu fraction of the synthetic sample was free of Gd interference

  2. 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.

  3. 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...

  4. 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.

  5. Geochemistry of dissolved and suspended loads of the Seine River, France: anthropogenic impact, carbonate and silicate weathering

    Science.gov (United States)

    Roy, S.; Gaillardet, J.; Allègre, C. J.

    1999-05-01

    This study focuses on the chemistry of the Seine river system, one of the major rivers in Europe, and constitutes the first geochemical investigation of both suspended and dissolved loads of this river. The Seine river drains a typical Mesozoic-Cenozoic sedimentary basin: the Paris basin, constituted of limestones mixed or interbedded with terrigenous sediments derived from the paleoreliefs bordering the Mesozoic and Cenozoic seas. In the context of quantifying the global influence of carbonate and silicate weathering on atmospheric CO 2 consumption, the Seine river offers the possibility of examining weathering rates in a flat sedimentary environment, under temperate climatic conditions. One of the major problems associated with the Seine river, as with many temperate rivers, is pollution. We propose, in this paper, 2 approaches in order to correct the dissolved load of the Seine river for anthropogenic inputs and to calculate weathering rates of carbonates and silicates. The first uses the dissolved load of rivers and tries to allocate the different solutes to different sources. A mixing model, based on elemental ratios, is established and solved by an inversion technique. The second approach consists in using the suspended load geochemistry. Under steady state conditions, we show that the geochemistry of suspended sediments makes it possible to estimate the amount of solutes released during the chemical weathering of silicates, and thus to calculate weathering rates of silicates. The total dissolved load of the Seine river at Paris can be decomposed into 2% of solutes derived from natural atmospheric sources, 7% derived from anthropogenic atmospheric sources, 6% derived from agriculture, 3% derived from communal inputs, and 82% of solutes derived from rock weathering. During high floods, the contribution of atmospheric and agriculture inputs predominates. The weathering rate of carbonates is estimated to be 48 t/km 2/yr (25 mm/1000 yr). Only 10% of carbonates

  6. Surface Complexation Modeling of Fluoride Adsorption by Soil and the Role of Dissolved Aluminum on Adsorption

    Science.gov (United States)

    Padhi, S.; Tokunaga, T.

    2017-12-01

    Adsorption of fluoride (F) on soil can control the mobility of F and subsequent contamination of groundwater. Hence, accurate evaluation of adsorption equilibrium is a prerequisite for understanding transport and fate of F in the subsurface. While there have been studies for the adsorption behavior of F with respect to single mineral constituents based on surface complexation models (SCM), F adsorption to natural soil in the presence of complexing agents needs much investigation. We evaluated the adsorption processes of F on a natural granitic soil from Tsukuba, Japan, as a function of initial F concentration, ionic strength, and initial pH. A SCM was developed to model F adsorption behavior. Four possible surface complexation reactions were postulated with and without including dissolved aluminum (Al) and Al-F complex sorption. Decrease in F adsorption with the increase in initial pH was observed in between the initial pH range of 4 to 9, and a decrease in the rate of the reduction of adsorbed F with respect to the increase in the initial pH was observed in the initial pH range of 5 to 7. Ionic strength variation in the range of 0 to 100mM had insignificant effect on F removal. Changes in solution pH were observed by comparing the solution before and after F adsorption experiments. At acidic pH, the solution pH increased, whereas at alkaline pH, the solution pH decreased after equilibrium. The SCM including dissolved Al and the adsorption of Al-F complex can simulate the experimental results quite successfully. Also, including dissolved Al and the adsorption of Al-F complex to the model explained the change in solution pH after F adsorption.

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

    International Nuclear Information System (INIS)

    Wang, R.

    1981-03-01

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

  8. Engineering study: Fast Flux Test Facility fuel reprocessing

    International Nuclear Information System (INIS)

    Beary, M.M.; Raab, G.J.; Reynolds, W.R. Jr.; Yoder, R.A.

    1974-01-01

    Several alternatives were studied for reprocessing FFTF fuels at Hanford. Alternative I would be to decontaminate and trim the fuel at T Plant and electrolytically dissolve the fuel at Purex. Alternative II would be to decontaminate and shear leach the fuels in a new facility near Purex. Alternative III would be to decontaminate and store fuel elements indefinitely at T Plant for subsequent offsite shipment. Alternative I, 8 to 10 M$ and 13 quarter-years; for Alternative II, 24 to 28 M$ and 20 quarter-years; for Alternative III, 3 to 4 M$ and 8 quarter-years. Unless there is considerable slippage in the FFTF shipping schedule, it would not be possible to build a new facility as described in Alternative II in time without building temporary storage facilities at T Plant, as described in Alternative III

  9. EFFECTIVENESS OF USING DILUTE OXALIC ACID TO DISSOLVE HIGH LEVEL WASTE IRON BASED SLUDGE SIMULANT

    International Nuclear Information System (INIS)

    Ketusky, E

    2008-01-01

    At the Savannah River Site (SRS), near Aiken South Carolina, there is a crucial need to remove residual quantities of highly radioactive iron-based sludge from large select underground storage tanks (e.g., 19,000 liters of sludge per tank), in order to support tank closure. The use of oxalic acid is planned to dissolve the residual sludge, hence, helping in the removal. Based on rigorous testing, primarily using 4 and 8 wt% oxalic acid solutions, it was concluded that the more concentrated the acid, the greater the amount of residual sludge that would be dissolved; hence, a baseline technology on using 8 wt% oxalic acid was developed. In stark contrast to the baseline technology, reports from other industries suggest that the dissolution will most effectively occur at 1 wt% oxalic acid (i.e., maintaining the pH near 2). The driver for using less oxalic acid is that less (i.e., moles) would decrease the severity of the downstream impacts (i.e., required oxalate solids removal efforts). To determine the initial feasibility of using 1 wt% acid to dissolve > 90% of the sludge solids, about 19,000 liters of representative sludge was modeled using about 530,000 liters of 0 to 8 wt% oxalic acid solutions. With the chemical thermodynamic equilibrium based software results showing that 1 wt% oxalic acid could theoretically work, simulant dissolution testing was initiated. For the dissolution testing, existing simulant was obtained, and an approximate 20 liter test rig was built. Multiple batch dissolutions of both wet and air-dried simulant were performed. Overall, the testing showed that dilute oxalic acid dissolved a greater fraction of the stimulant and resulted in a significantly larger acid effectiveness (i.e., grams of sludge dissolved/mole of acid) than the baseline technology. With the potential effectiveness confirmed via simulant testing, additional testing, including radioactive sludge testing, is planned

  10. 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

  11. A mathematical model and optimization of the cathode catalyst layer structure in PEM fuel cells

    International Nuclear Information System (INIS)

    Wang Qianpu; Song Datong; Navessin, Titichai; Holdcroft, Steven; Liu Zhongsheng

    2004-01-01

    A spherical flooded-agglomerate model for the cathode catalyst layer of a proton exchange membrane fuel cell, which includes the kinetics of oxygen reduction, at the catalyst vertical bar electrolyte interface, proton transport through the polymer electrolyte network, the oxygen diffusion through gas pore, and the dissolved oxygen diffusion through electrolyte, is considered. Analytical and numerical solutions are obtained in various control regimes. These are the limits of (i) oxygen diffusion control (ii) proton conductivity control, and (iii) mixture control. The structure and material parameters, such as porosity, agglomerate size, catalyst layer thickness and proton conductivity, on the performance are investigated under these limits. The model could help to characterize the system properties and operation modes, and to optimize catalyst layer design

  12. LIFE Materails: Molten-Salt Fuels Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R; Brown, N; Caro, A; Farmer, J; Halsey, W; Kaufman, L; Kramer, K; Latkowski, J; Powers, J; Shaw, H; Turchi, P

    2008-12-11

    The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

  13. LIFE Materails: Molten-Salt Fuels Volume 8

    International Nuclear Information System (INIS)

    Moir, R.; Brown, N.; Caro, A.; Farmer, J.; Halsey, W.; Kaufman, L.; Kramer, K.; Latkowski, J.; Powers, J.; Shaw, H.; Turchi, P.

    2008-01-01

    The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

  14. A study on the radioactive waste management for DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Kwan Sik; Park, H. S.; Park, J. J.; Kim, J. H.; Cho, Y. H.; Shin, J. M.; Kim, Y. K.; Kim, J. S.; Kim, J. G.; Park, S. D.; Suh, M. Y.; Sohn, S. C.; Song, B. C.; Lee, C. H.; Jeon, Y. S.; Jo, K. S.; Jee, K. Y.; Jee, C. S.; Han, S. H.

    1997-09-01

    Part 1: The characteristics if the radioactive wastes coming from the DUPIC fuel manufacturing process were analyzed and evaluated. The gross {alpha}-activity and {alpha}-, {gamma}-spectrum of irradiated zircaloy specimens form KORI unit 1 were analyzed. In order to develop the trapping media of radioactive ruthenium oxides, trapping behavior of volatilized ruthenium oxides on various metal oxides or carbonates was analyzed. Fly ash was selected as a trapping materials for gaseous cesium. And reaction characteristics of CsNO{sub 3} and CsI with fly ash have been investigated. Also, trapping material were performed to test fly ash filter for removal of gaseous cesium under the air and hydrogen atmosphere. The applicability of fly ash to the vitrification of the spent filter was analyzed in the aspects of predictability, leachability. Good quality of Borosilicate glass was formed using Cesium spent filter. Offgas treatment system of DUPIC fuel manufacturing facility was designed and constructed in order to trap of gaseous radioactive waste from 100 batch of OREOXA furnace (the capacity : 500 g/batch). Part II: To develop chemical analysis techniques necessary for understanding chemical properties of the highly radioactive materials related to the development of DUPIC fuel cycle technology, the following basic studies were performed : dissolution of SIMFUEL (simulated fuel), determination of uranium by potentiometry and UV/Vis absorption spectrophotometry, separation of PWR spent fuel, group separation of fission products from uranium, individual separation for analysis of actinides, determination of free acid in a artificial dissolved solution of PWR spent fuel, group separation of fission products form uranium, individual separation of Sm from a mixed rare earth elements and measurement of its isotopes by TI-mass spectrometry, and characteristics of detectors in inductively coupled plasma atomic emission spectrometer (ICP-AES) suitable for analysis of trace fission

  15. A study on the radioactive waste management for DUPIC fuel cycle

    International Nuclear Information System (INIS)

    Chun, Kwan Sik; Park, H. S.; Park, J. J.; Kim, J. H.; Cho, Y. H.; Shin, J. M.; Kim, Y. K.; Kim, J. S.; Kim, J. G.; Park, S. D.; Suh, M. Y.; Sohn, S. C.; Song, B. C.; Lee, C. H.; Jeon, Y. S.; Jo, K. S.; Jee, K. Y.; Jee, C. S.; Han, S. H.

    1997-09-01

    Part 1: The characteristics if the radioactive wastes coming from the DUPIC fuel manufacturing process were analyzed and evaluated. The gross α-activity and α-, γ-spectrum of irradiated zircaloy specimens form KORI unit 1 were analyzed. In order to develop the trapping media of radioactive ruthenium oxides, trapping behavior of volatilized ruthenium oxides on various metal oxides or carbonates was analyzed. Fly ash was selected as a trapping materials for gaseous cesium. And reaction characteristics of CsNO 3 and CsI with fly ash have been investigated. Also, trapping material were performed to test fly ash filter for removal of gaseous cesium under the air and hydrogen atmosphere. The applicability of fly ash to the vitrification of the spent filter was analyzed in the aspects of predictability, leachability. Good quality of Borosilicate glass was formed using Cesium spent filter. Offgas treatment system of DUPIC fuel manufacturing facility was designed and constructed in order to trap of gaseous radioactive waste from 100 batch of OREOXA furnace (the capacity : 500 g/batch). Part II: To develop chemical analysis techniques necessary for understanding chemical properties of the highly radioactive materials related to the development of DUPIC fuel cycle technology, the following basic studies were performed : dissolution of SIMFUEL (simulated fuel), determination of uranium by potentiometry and UV/Vis absorption spectrophotometry, separation of PWR spent fuel, group separation of fission products from uranium, individual separation for analysis of actinides, determination of free acid in a artificial dissolved solution of PWR spent fuel, group separation of fission products form uranium, individual separation of Sm from a mixed rare earth elements and measurement of its isotopes by TI-mass spectrometry, and characteristics of detectors in inductively coupled plasma atomic emission spectrometer (ICP-AES) suitable for analysis of trace fission products. (author

  16. 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

  17. General Atomic Reprocessing Pilot Plant: engineering-scale dissolution system description

    International Nuclear Information System (INIS)

    Yip, H.H.

    1979-04-01

    In February 1978, a dissolver-centrifuge system was added to the cold reprocessing pilot plant at General Atomic Company, which completed the installation of an HTGR fuel head-end reprocessing pilot plant. This report describes the engineering-scale equipment in the pilot plant and summarizes the design features derived from development work performed in the last few years. The dissolver operating cycles for both thorium containing BISO and uranium containinng WAR fissile fuels are included. A continuous vertical centrifuge is used to clarify the resultant dissolver product solution. Process instrumentation and controls for the system reflect design philosophy suitable for remote operation

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

    Science.gov (United States)

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

    2013-03-01

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

  19. 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.

  20. Flash pyrolysis of coal-solvent slurry prepared from the oxidized coal and the coal dissolved in solvent; Ichibu yokaishita sanka kaishitsutan slurry no jinsoku netsubunkai

    Energy Technology Data Exchange (ETDEWEB)

    Maki, T.; Mae, K.; Okutsu, H.; Miura, K. [Kyushu University, Fukuoka (Japan). Faculty of Engineering

    1996-10-28

    In order to develop a high-efficiency coal pyrolysis method, flash pyrolysis was experimented on slurry prepared by using liquid-phase oxidation reformed coal and a methanol-based solvent mixture. Australian Morwell coal was used for the experiment. The oxidized coal, into which carboxyl groups have been introduced, has the condensation structure relaxed largely, and becomes highly fluid slurry by means of the solvent. Char production can be suppressed by making the oxidation-pretreated coal into slurry, resulting in drastically improved pyrolytic conversion. The slurry was divided into dissolved solution, dried substance, extracted residue, and residual slurry, which were pyrolized independently. The dissolved solution showed very high conversion. Improvement in the conversion is contributed by separating the dissolved substances (coal macromolecules) at molecular levels, coagulating the molecules, suppressing cross-link formation, and reducing molecular weight of the dissolved substances. Oxidized coal can be dissolved to 80% or higher by using several kinds of mixed solvents. As a result of the dissolution, a possibility was suggested on pyrolysis which is easy in handling and high in conversion. 7 refs., 6 figs., 2 tabs.

  1. Measurement of dose rate and estimation of beta activity in zircaloy hull drum

    International Nuclear Information System (INIS)

    Pandey, J.P.N.; Kumar, Pankaj; Shinde, A.M.; Purohit, R.G.; Sarkar, P.K.

    2012-01-01

    Fuel Reprocessing Plant is designed for the processing of spent fuel from reactor for the recovery of plutonium and uranium as PuO 2 and U 3 O 8 respectively. Zircaloy is used as cladding material of natural uranium fuel pins used in the reactors. In reprocessing plants chop and leach method is used to remove the zircaloy clad from the fuel matrix during Head End Treatment. Initially spent fuel bundles are chopped into pieces and collected in perforated baskets kept in dissolvers. All chopped pieces are dissolved in HNO 3 in the dissolvers followed by heating and boiling. Dissolved solutions are transferred to Filtrate Tank (FT) leaving behind un-dissolved zircoloy hull pieces in the dissolver baskets. Un-dissolved and almost dry hull pieces are transferred in hull drum from the dissolver baskets using the Hull Tilting Facility. Hull drums are made of stainless steel having 500 litre capacity and two third of its volume is filled with zircoloy pieces. Hull drums filled with hull pieces are loaded in Hull Removal Cask (HRC) and transported to SWMF (Solid Waste Management Facility) site for interim storage/disposal in tile holes. Hull pieces are high active solid wastes which contain significant amount of fission products. Radiation levels on hull drums are in the range of few hundreds of mGy/h which has high potential of external hazards if not handled properly. Therefore hull drums are handled remotely in specially designed lead shielded cask

  2. 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

  3. Fuel particles in the Chernobyl cooling pond: current state and prediction for remediation options

    International Nuclear Information System (INIS)

    Bulgakov, A.; Konoplev, A.; Smith, J.; Laptev, G.; Voitsekhovich, O.

    2009-01-01

    During the coming years, a management and remediation strategy for the Chernobyl cooling pond (CP) will be implemented. Remediation options include a controlled reduction in surface water level of the cooling pond and stabilisation of exposed sediments. In terrestrial soils, fuel particles deposited during the Chernobyl accident have now almost completely disintegrated. However, in the CP sediments the majority of 90 Sr activity is still in the form of fuel particles. Due to the low dissolved oxygen concentration and high pH, dissolution of fuel particles in the CP sediments is significantly slower than in soils. After the planned cessation of water pumping from the Pripyat River to the Pond, significant areas of sediments will be drained and exposed to the air. This will significantly enhance the dissolution rate and, correspondingly, the mobility and bioavailability of radionuclides will increase with time. The rate of acidification of exposed bottom sediments was predicted on the basis of acidification of similar soils after liming. Using empirical equations relating the fuel particle dissolution rate to soil and sediment pH allowed prediction of fuel particle dissolution and 90 Sr mobilisation for different remediation scenarios. It is shown that in exposed sediments, fuel particles will be almost completely dissolved in 15-25 years, while in parts of the cooling pond which remain flooded, fuel particle dissolution will take about a century

  4. 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

  5. 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.

  6. Formation and properties of radiocolloids in aqueous solution - a literature survey

    International Nuclear Information System (INIS)

    Olofsson, U.; Allard, B.; Andersson, K.; Torstenfelt, B.

    1981-06-01

    The sorption of radionuclides on various rocks and minerals has been studied within many national waste programs as a means of predicting the migration behaviour of radionuclides that might be released from e.g. an underground repository for radioactive waste. One major objection against the conclusions that can be drawn from laboratory sorption studies is that the possibility of a formation of small fractions of highly mobile particulates are usually not considered. The elements, present in spent nuclear fuel, which are most likely to form colloid species would be hydrolyzable elements like the actinides and possibly Sr as well as Pb and Cu representing the encapsulation material. Moreover the radionuclides would be present in aqueous solutions in very low concentrations and under these conditions other phenomena occurs than at macroconcentrations. This literature survey is meant to be a basis for further studies on the formation and transport of radiocolloids in the groundwater-rock environment. The colloids will probably not be retarded by the same mechanisms as dissolved species in true solution, but may in some cases migrate with the same velocity as the groundwater. (Auth.)

  7. Dissolution study of thorium-uranium oxides in aqueous triflic acid solutions

    Energy Technology Data Exchange (ETDEWEB)

    Bulemela, E.; Bergeron, A.; Stoddard, T. [Canadian Nuclear Laboratories - CNL, 286 Plant Rd., Chalk River, Ontario, K0J 1J0 (Canada)

    2016-07-01

    The dissolution of sintered mixed oxides of thorium with uranium in various concentrations of trifluoromethanesulfonic (triflic) acid solutions was investigated under reflux conditions to evaluate the suitability of the method. Various fragment sizes (1.00 mm < x < 7.30 mm) of sintered (Th,U)O{sub 2} and simulated high-burnup nuclear fuel (SIMFUEL) were almost completely dissolved in a few hours, which implies that triflic acid could be used as an alternative to the common dissolution method, involving nitric acid-hydrofluoric acid mixture. The influence of acid concentration, composition of the solids, and reaction time on the dissolution yield of Th and U ions was studied using Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). The dissolution rate was found to depend upon the triflic acid concentration and size of the solid fragments, with near complete dissolution for the smallest fragments occurring in boiling 87% w/w triflic acid. The formation of Th and U ions in solution appears to occur at the same rate as the triflic acid simultaneously reacts with the constituent oxides as evidenced by the results of a constant U/Th concentration ratio with the progress of the dissolution. (authors)

  8. 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)

  9. Evaluation on corrosively dissolved gold induced by alkanethiol monolayer with atomic absorption spectroscopy

    International Nuclear Information System (INIS)

    Cao Zhong; Zhang Ling; Guo Chaoyan; Gong Fuchun; Long Shu; Tan Shuzhen; Xia Changbin; Xu Fen; Sun Lixian

    2009-01-01

    We have monitored a gold corrosive dissolution behavior accompanied in n-alkanethiol like n-dodecanethiol assembled process with in situ quartz crystal microbalance (QCM), and then observed it with atomic force microscopy (AFM) which showed an evident image of corrosive defects or holes produced on gold substrate, corresponding to gold dissolution induced by the alkanethiol molecules in the presence of oxygen. For detection of the dissolved gold defects during alkanethiol assembled process, an atomic absorption spectroscopy (AAS) has been carried out in this paper, and the detection limit for the dissolved gold could be evaluated to be 15.4 ng/mL. The amount of dissolved gold from the substrates of gold plates as functions of immersion time, acid media, solvents and thiol concentration has been examined in the oxygen saturated solutions. In comparison with in situ QCM method, the kinetics behavior of the long-term gold corrosion on the gold plates in 1.0 mmol/L of n-dodecanethiol solution determined with AAS method was a slow process, and its corrosion rate on gold dissolution could be evaluated to be about 4.4 x 10 -5 ng.cm -2 .s -1 , corresponding to 1.3 x 10 8 Au atoms.cm -2 .s -1 , that was much smaller than that of initial rate monitored with in situ QCM. Both kinetics equations obtained with QCM and AAS showed a consistent corrosion behavior on gold surfaces.

  10. 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)

  11. Recovery of fission products from acidic waste solutions thereof

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.; Dubois, D.W.

    1975-01-01

    Fission products, e.g., palladium, ruthenium and technetium, are removed from aqueous, acidic waste solutions thereof. The acidic waste solution is electrolyzed in an electrolytic cell under controlled cathodic potential conditions and technetium, ruthenium, palladium and rhodium are deposited on the cathode. Metal deposit is removed from the cathode and dissolved in acid. Acid insoluble rhodium metal is recovered, dissolved by alkali metal bisulfate fusion and purified by electrolysis. In one embodiment, the solution formed by acid dissolution of the cathode metal deposit is treated with a strong oxidizing agent and distilled to separate technetium and ruthenium (as a distillate) from palladium. Technetium is separated from ruthenium by organic solvent extraction and then recovered, e.g., as an ammonium salt. Ruthenium is disposed of as waste by-product. Palladium is recovered by electrolysis of an acid solution thereof under controlled cathodic potential conditions. Further embodiments wherein alternate metal recovery sequences are used are described. (U.S.)

  12. Development of isotope dilution gamma-ray spectrometry for plutonium analysis

    Energy Technology Data Exchange (ETDEWEB)

    Li, T.K.; Parker, J.L. (Los Alamos National Lab., NM (United States)); Kuno, Y.; Sato, S.; Kurosawa, A.; Akiyama, T. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))

    1991-01-01

    We are studying the feasibility of determining the plutonium concentration and isotopic distribution of highly radioactive, spent-fuel dissolver solutions by employing high-resolution gamma-ray spectrometry. The study involves gamma-ray plutonium isotopic analysis for both dissolver and spiked dissolver solution samples, after plutonium is eluted through an ion-exchange column and absorbed in a small resin bead bag. The spike is well characterized, dry plutonium containing {approximately}98% of {sup 239}Pu. By using measured isotopic information, the concentration of elemental plutonium in the dissolver solution can be determined. Both the plutonium concentration and the isotopic composition of the dissolver solution obtained from this study agree well with values obtained by traditional isotope dilution mass spectrometry (IDMS). Because it is rapid, easy to operate and maintain, and costs less, this new technique could be an alternative method to IDMS for input accountability and verification measurements in reprocessing plants. 7 refs., 4 figs., 4 tabs.

  13. Volumetric properties of itaconic acid aqueous solutions

    International Nuclear Information System (INIS)

    Nisenbaum, Alexander; Apelblat, Alexander; Manzurola, Emanuel

    2012-01-01

    Highlights: ► Densities of itaconic acid aqueous solutions in a wide range of concentrations and temperatures. ► The apparent molar volumes and the cubic expansion coefficients. ► The derivatives of isobaric heat capacities with respect to pressure. ► Changes in the structure of water when itaconic acid is dissolved. - Abstract: Densities of itaconic acid aqueous solutions were measured at 5 K intervals from T = (278.15 to 343.15) K. From the determined densities, the apparent molar volumes, the cubic expansion coefficients and the second derivatives of volume with respect to temperature which are interrelated with the derivatives of isobaric heat capacities with respect to pressure were evaluated. These derivatives were qualitatively correlated with the changes in the structure of water when itaconic acid is dissolved in it.

  14. The effect of the oxygen dissolved in the adsorption of gold in activated carbon; Efecto del oxigeno disuelto en la adsorcion de oro en carbon activado

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, P. [Universidad de Santiago. Chile (Chile); Wilkomirsky, I. [Universidad de Concepcion. Chile (Chile)

    1999-07-01

    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.

  15. 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

  16. Hazards classification determination for PUREX fuel transfer to K-Basins

    International Nuclear Information System (INIS)

    Dodd, E.N. III.

    1995-01-01

    The PUREX Plant presently contains 2.9 metric tons of an aluminum clad Single Pass Reactor (SPR) fuel which is stored under water in four open top buckets in the PUREX slug storage basin. The PUREX dissolver cells contain approximately 0.5 metric tons of zirconium clad N Reactor fuel which was inadvertently placed into the process cell during charging operations. The dissolver N reactor elements will be recovered from the process floors using new crane operated tools. When the fuel shipment(s) is scheduled, the cask cars will be positioned into the PUREX rail tunnel and the overhead door will be opened. All the SPR fuel will be loaded into two cask rail cars inside four casks. The N Reactor fuel will be loaded into a separate rail car inside two or three casks. The car loading is initiated by opening the rail car lid and removing the cask lids. Prior to loading the canisters of N Reactor fuel, the canisters will be refilled with water (as needed) and a lid will be installed. The baskets of SPR fuel or canisters of N Reactor fuel will then be loaded into the casks. The lids to the casks will then be reinstalled and the car lids closed. The rail cars will then be decontaminated as necessary. The cask cars will be shipped either in two shipments or a combined single shipment using the rail route between PUREX and the K Basins. At the basin, the cask car will be positioned in the loadout area. The cask car lid will be opened and a single cask moved into the loadout pit, which is a lowered section of the basin. The cask lid is removed while the cask is lower into the pit. The fuel is then removed from the cask and stored in the basin. The cask is then removed, the lid reinstalled during removal, and the cask replaced into the cask car. This document identifies the hazard classification of the Fuel Transfer from the PUREX facility to K-Basins

  17. Dynamic leaching studies of 48 MWd/kgU UO{sub 2} commercial spent nuclear fuel under oxic conditions

    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, Karlsruhe (Germany); Casas, I. [Department of Chemical Engineering, UPC, Barcelona (Spain); González-Robles, E. [Environmental Technology Department, Fundació CTM Centre Tecnològic, Manresa, Barcelona (Spain); Glatz, J.P.; Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe (Germany); Clarens, F. [Environmental Technology Department, Fundació CTM Centre Tecnològic, Manresa, Barcelona (Spain); Giménez, J. [Department of Chemical Engineering, UPC, Barcelona (Spain); Pablo, J. de [Department of Chemical Engineering, UPC, Barcelona (Spain); Environmental Technology Department, Fundació CTM Centre Tecnològic, Manresa, Barcelona (Spain); Martínez-Esparza, A. [High Level Waste Department, ENRESA, Empresa Nacional de Residuos Radioactivos, Madrid (Spain)

    2013-03-15

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

  18. The characterization of insoluble dissolver residues and the development of treatment methods

    International Nuclear Information System (INIS)

    Baker, H.T.; Brown, P.E.; Pateman, R.J.; Wilkinson, K.L.

    1986-01-01

    Characterization studies have been carried out on the insoluble residue arising from laboratory scale dissolution of single pins of mixed oxide fuel irradiated in the Dounreay Fast Reactor (DFR). Similar characterization work has also been completed on six samples of insoluble residue recovered from the reprocessing of complete sub-assemblies of mixed oxide fuel irradiated in PFR. Treatment methods for the dissolver insolubles have consisted of preparing immobilized samples in sand/ordinary portland cement (OPC), sand/pulverized fly ash (PFA)/OPC, and blast furnace slag (BFS)/OPC. A programme of leach testing has been carried out according to the conditions laid down in the international Standard Organization Test. Four samples of DFR originated insoluble residues and six samples of PFR originated insoluble residues have been immobilized and leach tested. Variations have included experiments to evaluate the leach rate under temperature influence and to acid wash prior to immobilization

  19. 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.

  20. 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.

  1. 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).

  2. 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)

  3. New phenomena observed during fuel assemblies testing

    International Nuclear Information System (INIS)

    Tzotcheva, V.

    2001-01-01

    The paper presents a new attempt to explain inexplicable increase of specific activity for some of the fuel assemblies during the fuel tightness testing procedures on Kozloduy NPP. A brief description of established procedure for fuel tightness control is presented in the paper. Special emphasis is given on a hypothesis that explains the fact of existence of deviation in Iodine activity more than usual, which have no reasonable interpretation. The reasons for uniform high Iodine activity for reloaded assemblies, that have kept in the open measuring can for a long time (1-3 hours), is found to be the process of Iodine dissolving in the water and the accelerated process of natural degassing. A proposal to use the 134 Cs and 137 Cs as stand-alone criteria for more precise results is made in respect to increase the reliability of fuel reloading and storage procedures

  4. The soil organic carbon content of anthropogenically altered organic soils effects the dissolved organic matter quality, but not the dissolved organic carbon concentrations

    Science.gov (United States)

    Frank, Stefan; Tiemeyer, Bärbel; Bechtold, Michel; Lücke, Andreas; Bol, Roland

    2016-04-01

    Dissolved organic carbon (DOC) is an important link between terrestrial and aquatic ecosystems. This is especially true for peatlands which usually show high concentrations of DOC due to the high stocks of soil organic carbon (SOC). Most previous studies found that DOC concentrations in the soil solution depend on the SOC content. Thus, one would expect low DOC concentrations in peatlands which have anthropogenically been altered by mixing with sand. Here, we want to show the effect of SOC and groundwater level on the quantity and quality of the dissolved organic matter (DOM). Three sampling sites were installed in a strongly disturbed bog. Two sites differ in SOC (Site A: 48%, Site B: 9%) but show the same mean annual groundwater level of 15 and 18 cm below ground, respectively. The SOC content of site C (11%) is similar to Site B, but the groundwater level is much lower (-31 cm) than at the other two sites. All sites have a similar depth of the organic horizon (30 cm) and the same land-use (low-intensity sheep grazing). Over two years, the soil solution was sampled bi-weekly in three depths (15, 30 and 60 cm) and three replicates. All samples were analyzed for DOC and selected samples for dissolved organic nitrogen (DON) and delta-13C and delta-15N. Despite differences in SOC and groundwater level, DOC concentrations did not differ significantly (A: 192 ± 62 mg/L, B: 163 ± 55 mg/L and C: 191 ± 97 mg/L). At all sites, DOC concentrations exceed typical values for peatlands by far and emphasize the relevance even of strongly disturbed organic soils for DOC losses. Individual DOC concentrations were controlled by the temperature and the groundwater level over the preceding weeks. Differences in DOM quality were clearer. At site B with a low SOC content, the DOC:DON ratio of the soil solution equals the soil's C:N ratio, but the DOC:DON ratio is much higher than the C:N ratio at site A. In all cases, the DOC:DON ratio strongly correlates with delta-13C. There is no

  5. Changing fluxes of carbon and other solutes from the Mekong River.

    Science.gov (United States)

    Li, Siyue; Bush, Richard T

    2015-11-02

    Rivers are an important aquatic conduit that connects terrestrial sources of dissolved inorganic carbon (DIC) and other elements with oceanic reservoirs. The Mekong River, one of the world's largest rivers, is firstly examined to explore inter-annual fluxes of dissolved and particulate constituents during 1923-2011 and their associated natural or anthropogenic controls. Over this period, inter-annual fluxes of dissolved and particulate constituents decrease, while anthropogenic activities have doubled the relative abundance of SO4(2-), Cl(-) and Na(+). The estimated fluxes of solutes from the Mekong decrease as follows (Mt/y): TDS (40.4) > HCO3(-) (23.4) > Ca(2+) (6.4) > SO4(2-) (3.8) > Cl(-) (1.74)~Na(+) (1.7) ~ Si (1.67) > Mg(2+) (1.2) > K(+ 0.5). The runoff, land cover and lithological composition significantly contribute to dissolved and particulate yields globally. HCO3(-) and TDS yields are readily predicted by runoff and percent of carbonate, while TSS yield by runoff and population density. The Himalayan Rivers, including the Mekong, are a disproportionally high contributor to global riverine carbon and other solute budgets, and are of course underlined. The estimated global riverine HCO3(-) flux (Himalayan Rivers included) is 34,014 × 10(9) mol/y (0.41 Pg C/y), 3915 Mt/y for solute load, including HCO3(-), and 13,553 Mt/y for TSS. Thereby this study illustrates the importance of riverine solute delivery in global carbon cycling.

  6. Process for the manufacture of a fuel catalyst made of tungsten carbide for electrochemical fuel cells. Verfahren zur Herstellung eines Brennstoffkatalysators aus Wolframcarbid fuer elektrochemische Brennstoffzellen

    Energy Technology Data Exchange (ETDEWEB)

    Baresel, D.; Gellert, W.; Scharner, P.

    1982-05-19

    The invention refers to a process for the manufacture of a fuel catalyst made of tungsten carbide for the direct generation of electrical energy by the oxidation of hydrogen, formaldehyde or formic acid in electrochemical fuel cells. Tungsten carbide is obtained by carburisation of tungsten or tungsten oxide by carbon monoxide. The steps of the process are as follows: dissolving the commercial-quality tungstic acid in ammonium hydroxide; precipitating the tungstic acid with concentrated hydrochloric acid; drying in a vacuum and then heating to 200/sup 0/C to remove the water of crystallisation forming tungsten trioxide; and mixing the tungsten trioxide with zinc powder and heating to 600/sup 0/C. The zinc oxide is dissolved with hydrochloric acid after cooling. The finely divided tungsten obtained in this way is converted with carbon monoxide in a quartz tube at 700/sup 0/C.

  7. The reducibility of sulphuric acid and sulphate in aqueous solution (translated from German)

    International Nuclear Information System (INIS)

    Grauer, R.

    1990-07-01

    In connection with the Swedish project for the final storage of spent fuel elements it was necessary to assess whether dissolved sulphate can corrode the copper canister without the intervention of sulphate-reducing bacteria. A simple reaction between copper and sulphate is thermodynamically impossible. On the other hand, copper can react to give copper sulphide if an additional electron donor such as iron is available. Because little specific information is available about this subject the problem was extended to the much more general question of the reducibility of sulphur in dilute aqueous solution. It is a part of the general knowledge of chemistry, and there is also unanimity about it in the geochemical literature, that purely chemical reduction of sulphate does not take place in dilute solution at temperatures below 100 degrees C. This fact is, however, poorly documented and it was therefore necessary to substantiate it by drawing on numerous individual findings from different areas of pure and applied chemistry. The investigation confirms that sulphur in dilute solution is completely inert towards chemical reducing agents and also to cathodic reduction. Thus corrosion of copper by sulphate under final-storage conditions and in the absence of sulphate reducing bacteria can be ruled out with a probability verging on certainty. (85 refs.)

  8. 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)

  9. Transport of Liquid Phase Organic Solutes in Liquid Crystalline Membranes

    OpenAIRE

    Han, Sangil

    2010-01-01

    Porous cellulose nitrate membranes were impregnated with 8CB and PCH5 LCs (liquid crystals) and separations of solutes dissolved in aqueous phases were performed while monitoring solute concentration via UV-VIS spectrometry. The diffusing organic solutes, which consist of one aromatic ring and various functional groups, were selected to exclude molecular size effects on the diffusion and sorption. We studied the effects on solute transport of solute intra-molecular hydrogen bonding and so...

  10. Assessing the effect of dissolved organic ligands on mineral dissolution rates: An example from calcite dissolution

    International Nuclear Information System (INIS)

    DeMaio, T.; Grandstaff, D.E.

    1997-01-01

    Experiments suggest that dissolved organic ligands may primarily modify mineral dissolution rates by three mechanisms: (1) metal-ligand (M-L) complex formation in solution, which increases the degree of undersaturation, (2) formation of surface M-L complexes that attack the surface, and (3) formation of surface complexes which passivate or protect the surface. Mechanisms (1) and (2) increase the dissolution rate and the third decreases it compared with organic-free solutions. The types and importance of these mechanisms may be assessed from plots of dissolution rate versus degree of undersaturation. To illustrate this technique, calcite, a common repository cementing and vein-filling mineral, was dissolved at pH 7.8 and 22 C in Na-Ca-HCO 3 -Cl solutions with low concentrations of three organic ligands. Low citrate concentrations (50 microM) increased the dissolution rate consistent with mechanism (1). Oxalate decreased the rate, consistent with mechanism (3). Low phthalate concentration (<50 microM) decreased calcite dissolution rates; however, higher concentrations increased the dissolution rates, which became faster than in inorganic solutions. Thus, phthalate exhibits both mechanisms (2) and (3) at different concentrations. In such cases linear extrapolations of dissolution rates from high organic ligand concentrations may not be valid

  11. Outline of a fuel treatment facility in NUCEF

    International Nuclear Information System (INIS)

    Sugikawa, Susumu; Umeda, Miki; Kokusen, Junya

    1997-03-01

    This report presents outline of the nuclear fuel treatment facility for the purpose of preparing solution fuel used in Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) in Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF), including descriptions of process conditions and dimensions of major process equipments on dissolution system of oxide fuel, chemical adjustment system, purification system, acid recovery system, solution fuel storage system, and descriptions of safety design philosophy such as safety considerations of criticality, solvent fire, explosion of hydrogen and red-oil and so on. (author)

  12. 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

  13. Role of natural dissolved organic compounds in determining the concentrations of americium in natural waters

    International Nuclear Information System (INIS)

    Nelson, D.M.; Orlandini, K.A.

    1985-01-01

    Concentrations of 241 Am, both in solution and bound to suspended particulate matter, have been measured in several North American lakes. Dissolved concentrations vary from 0.4 μBq/L to 85 μBq/L. The 241 Am in these lakes originated solely from global fallout and hence entered all lakes in the same physiocochemical form. The observed differences in solubility behavior must, therefore, be attributable to chemical and/or hydrological differences among the lakes. Concentrations of dissolved 241 Am are highly correlated with the corresponding concentrations of /sup 239, 240/Pu(III,IV), suggesting that a common factor is responsible for maintaining both in solution. The K/sub D/ values for 241 Am and /sup 239, 240/Pu(III,IV) are highly correlated with the concentrations of dissolved organic carbon (DOC) in the waters, suggesting that the common factor is the formation of soluble complexes with natural DOC for both elements. This hypothesis was tested in a series of laboratory experiments in which the DOC from several of the lakes was isolated by ultrafiltration. Plots of K/sub D/, as a function of DOC concentration, show K/sub D/ to be very high (approx.10 6 ) at low DOC concentrations. Above critical concentrations (a few mg/L DOC) the K/sub D/ values begin a progressive decrease with increasing DOC. We conclude that in most surface waters, the dissolved 241 Am concentration is regulated by an adsorption/desorption equilibrium with the sediments (and suspended solids) and the value of K/sub D/ that characterizes this equilibrium is largely determined by the concentration of natural DOC in the water. 11 refs., 3 figs., 2 tabs

  14. From waste to traffic fuel (W-fuel)

    Energy Technology Data Exchange (ETDEWEB)

    Kask, Ue.; Andrijevskaja, J.; Kask, L. [and others

    2012-11-01

    The EU directive on the promotion of the use of energy from renewable sources (Directive 2009/28/EC) sets a mandatory minimum target for the use of fuels produced using renewable energy sources of 10% of total petrol and diesel consumption in the transport sector by the year 2020. In addition, it states that production of renewable fuels should be consistent with sustainable development and must not endanger biodiversity. In the INTERREG IVA Southern Finland - Estonia Sub-programme, efforts towards finding solutions to the tasks set by the EU were undertaken in co-operation with Finnish and Estonian researchers. The purpose of the 'From Waste to Traffic Fuel' (W-Fuel) project was to promote the sustainable production and use of biogas using locally-sourced biodegradable waste materials from the food and beverage industry and the agricultural and municipal sectors. The ultimate aim of the project was to upgrade the biogas (produced based on anaerobic digestion of biodegradable wastes, sludge, manure, slurry and energy crops) to biomethane with a methane content similar to natural gas, to be further used as transport fuel with the aim of reducing traffic-borne emissions, in particular CO{sub 2}. The project combined waste, energy and traffic solutions in order to decrease emissions, costs and the use of materials. Six case areas in southern Finland and northern Estonia were selected. The two case areas in Estonia were the counties of Harju and Laeaene-Viru in northern Estonia. The project aimed to promote waste and sludge prevention and to commence biogas production and its subsequent upgrading to biomethane for use as a renewable fuel. The project promoted regional businesses and employment in waste treatment and 'green energy' production. On basis of the gathered data, the biogas potentials and prerequisites of each case county were analysed. Furthermore, the environmental, economic and other regional effects of the different options were

  15. HB-Line Dissolver Dilution Flows and Dissolution Capability with Dissolver Charge Chute Cover Off

    International Nuclear Information System (INIS)

    Hallman, D.F.

    2003-01-01

    A flow test was performed in Scrap Recovery of HB-Line to document the flow available for hydrogen dilution in the dissolvers when the charge chute covers are removed. Air flow through the dissolver charge chutes, with the covers off, was measured. A conservative estimate of experimental uncertainty was subtracted from the results. After subtraction, the test showed that there is 20 cubic feet per minute (cfm) air flow through the dissolvers during dissolution with a glovebox exhaust fan operating, even with the scrubber not operating. This test also showed there is 6.6 cfm air flow through the dissolvers, after subtraction of experimental uncertainty if the scrubber and the glovebox exhaust fans are not operating. Three H-Canyon exhaust fans provide sufficient motive force to give this 6.6 cfm flow. Material charged to the dissolver will be limited to chemical hydrogen generation rates that will be greater than or equal to 25 percent of the Lower Flammability Limit (LFL) during normal operations. The H-Canyon fans will maintain hydrogen below LFL if electrical power is lost. No modifications are needed in HB-Line Scrap Recovery to ensure hydrogen is maintained less that LFL if the scrubber and glovebox exhaust fans are not operating

  16. Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.

    1999-01-01

    Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

  17. 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

  18. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    International Nuclear Information System (INIS)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

    2015-01-01

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

  19. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  20. In-situ measurement of the dissolved S2- in seafloor diffuse flow system: sensor preparation and calibration

    Institute of Scientific and Technical Information of China (English)

    Ying YE; Xia HUANG; Yi-wen PAN; Chen-hua HAN; Wei ZHAO

    2008-01-01

    The preparation approach and calibration result of an improved type of ion selective electrode (ISE), which is used to measure the total dissolved S2-, are introduced in this paper. The improved Ag/Ag2S electrode uses silver wire as the substrate, which is surrounded by electric polymer containing superfine silver powder. After the stabilization of the epoxy-resin, Ag2ES layer was formed by chemical reaction with 0.2 mol/L (NH4)2S solution for 5 min. With Ag/AgCl as reference electrode, the Ag/Ag2S electrode can be used to measure dissolved S2-. The correlation between the measured potentials and the logarithm of dissolved S2- is found to be linear, within range of the concentration of dissolved S2- from 10-2~10-7 mol/L. The slope of the regression line between measured potential and logarithm of dissolved S2- is about -27.7, which agrees well with the theoretical Nernst value -29.6. Furthermore, the performance of the improved Ag/Ag2S electrode, such as the response time, sensitivity and stability, greatly outweighs the conventional Ag/Ag2S electrode.

  1. 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

  2. Effects of solution heat treatment on the microstructure and hardness of Mg-5Li-3Al-2Zn-2Cu alloy

    Energy Technology Data Exchange (ETDEWEB)

    Li Jiqing; An Jiangmin; Qu Zhikun [Key Laboratory of Superlight Materials and Surface Technology (Harbin Engineering University), Ministry of Education, Harbin 150001 (China); Wu Ruizhi, E-mail: Ruizhiwu2006@yahoo.com [Key Laboratory of Superlight Materials and Surface Technology (Harbin Engineering University), Ministry of Education, Harbin 150001 (China); Zhang Jinghuai; Zhang Milin [Key Laboratory of Superlight Materials and Surface Technology (Harbin Engineering University), Ministry of Education, Harbin 150001 (China)

    2010-10-15

    The microstructure and hardness of Mg-5Li-3Al-2Zn-2Cu alloy were investigated both in the as-cast condition and after solution heat treatment at 330-390 deg. C for 5 h. The as-cast alloy contains a microstructure consisting of {alpha}-Mg matrix, AlLi phase, AlCuMg phase and Al{sub 2}Cu phase. After the solution heat treament, the AlLi phase was dissolved into the matrix, however, the AlCuMg and Al{sub 2}Cu phases were not dissolved. With the increase of solution temperature, almost all the AlLi phase was dissolved, and the effects of solution strengthening of Al and Li atoms in the alloy increase, which results in the gradual increase of the Brinell hardness of the solution-treated alloy.

  3. Suppressive effects of a polymer sodium silicate solution on ...

    African Journals Online (AJOL)

    Sodium silicate was dissolved in water in either a monomer form or polymer form; the effects of both forms of sodium silicate aqueous solution on rose powdery mildew and root rot diseases of miniature rose were examined. Both forms of sodium silicate aqueous solution were applied to the roots of the miniature rose.

  4. Recovery of uranium from sulphate solutions containing molybdenum

    International Nuclear Information System (INIS)

    Weir, D.R.; Genik-Sas-Berezowsky, R.M.

    1983-01-01

    A process for recovering uranium from a sulphate solution containing dissolved uranium and molybdenum includes reacting the solution with ammonia (pH 8 to 10), the pH of the original solution must not exceed 5.5 and after the addition of ammonia the pH must not be in the vicinity of 7 for a significant time. The resultant uranium precipitate is relatively uncontaminated by molybdenum. The precipitate is then separated from the remaining solution while the pH is maintained within the stated range

  5. Fuel conditioning facility electrorefiner cadmium vapor trap operation

    International Nuclear Information System (INIS)

    Vaden, D. E.

    1998-01-01

    Processing sodium-bonded spent nuclear fuel at the Fuel Conditioning Facility at Argonne National Laboratory-West involves an electrometallurgical process employing a molten LiCl-KCl salt covering a pool of molten cadmium. Previous research has shown that the cadmium dissolves in the salt as a gas, diffuses through the salt layer and vaporizes at the salt surface. This cadmium vapor condenses on cool surfaces, causing equipment operation and handling problems. Using a cadmium vapor trap to condense the cadmium vapors and reflux them back to the electrorefiner has mitigated equipment problems and improved electrorefiner operations

  6. Changes in concentration and (delta) 13C value of dissolved CH4, CO2 and organic carbon in rice paddies under ambient and elevated concentrations of atmospheric CO2

    International Nuclear Information System (INIS)

    Weiguo Cheng; Yagi, Kazuyuki; Sakai, Hidemitsu; Hua Xu; Kobayashi, Kazuhiko

    2005-01-01

    Changes in concentration and (delta) 13 C value of dissolved CH 4 , CO 2 and organic carbon (DOC) in floodwater and soil solution from a Japanese rice paddy were studied under ambient and elevated concentrations of atmospheric CO 2 in controlled environment chambers. The concentrations of dissolved CH 4 in floodwater increased with rice growth (with some fluctuation), while the concentrations of CO 2 remained between 2.9 to 4.4 and 4.2 to 5.8 μg C mL -1 under conditions of ambient and elevated CO 2 concentration, respectively. The amount of CH 4 dissolved in soil solution under elevated CO 2 levels was significantly lower than under ambient CO 2 in the tillering stage, implying that the elevated CO 2 treatment accelerated CH 4 oxidation during the early stage of growth. However, during later stages of growth, production of CH 4 increased and the amount of CH 4 dissolved in soil solution under elevated CO 2 levels was, on average, greater than that under ambient CO 2 conditions. Significant correlation existed among the (delta) 13 C values of dissolved CH 4 , CO 2 , and DOC in floodwater (except for the samples taken immediately after pulse feeding with 13 C enriched CO 2 ), indicating that the origins and cycling of CH 4 , CO 2 and DOC were related. There were also significant correlations among the (delta) 13 C values of CH 4 , CO 2 and DOC in the soil solution. The turnover rate of CO 2 in soil solution was most rapid in the panicle formation stage of rice growth and that of CH 4 fastest in the grain filling stage. (Author)

  7. 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.)

  8. Foam-based adsorbents having high adsorption capacities for recovering dissolved metals and methods thereof

    Science.gov (United States)

    Janke, Christopher J.; Dai, Sheng; Oyola, Yatsandra

    2015-06-02

    Foam-based adsorbents and a related method of manufacture are provided. The foam-based adsorbents include polymer foam with grafted side chains and an increased surface area per unit weight to increase the adsorption of dissolved metals, for example uranium, from aqueous solutions. A method for forming the foam-based adsorbents includes irradiating polymer foam, grafting with polymerizable reactive monomers, reacting with hydroxylamine, and conditioning with an alkaline solution. Foam-based adsorbents formed according to the present method demonstrated a significantly improved uranium adsorption capacity per unit weight over existing adsorbents.

  9. 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)

  10. 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.

  11. Study on the Conversion of Fuel Nitrogen Into NOx

    Directory of Open Access Journals (Sweden)

    Raminta Plečkaitienė

    2011-12-01

    Full Text Available The aim of this work is to investigate NOx regularities combusting fuels having high concentration of nitrogen and to develop methods that will reduce the conversion of fuel nitrogen into NOx. There are three solutions to reducing NOx concentration: the combustion of fuel mixing it with other types of “clean” fuel containing small amounts of nitrogen, laundering fuel and the combustion of fuel using carbon additives. These solutions can help with reducing the amount of nitrogen in the wood waste of furniture by about 30% by washing fuel with water. Therefore, NOx value may decrease by about 35%.Article in Lithuanian

  12. Empirical evidence that soil carbon formation from plant inputs is positively related to microbial growth

    Science.gov (United States)

    Mark A. Bradford; Ashley D. Keiser; Christian A. Davies; Calley A. Mersmann; Michael S. Strickland

    2012-01-01

    Plant-carbon inputs to soils in the form of dissolved sugars, organic acids and amino acids fuel much of heterotrophic microbial activity belowground. Initial residence times of these compounds in the soil solution are on the order of hours, with microbial uptake a primary removal mechanism. Through microbial biosynthesis, the dissolved compounds become dominant...

  13. 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)

  14. Sediment-water interactions affecting dissolved-mercury distributions in Camp Far West Reservoir, California

    Science.gov (United States)

    Kuwabara, James S.; Alpers, Charles N.; Marvin-DiPasquale, Mark; Topping, Brent R.; Carter, James L.; Stewart, A. Robin; Fend, Steven V.; Parcheso, Francis; Moon, Gerald E.; Krabbenhoft, David P.

    2003-01-01

    Field and laboratory studies were conducted in April and November 2002 to provide the first direct measurements of the benthic flux of dissolved (0.2-micrometer filtered) mercury species (total and methylated forms) between the bottom sediment and water column at three sampling locations within Camp Far West Reservoir, California: one near the Bear River inlet to the reservoir, a second at a mid-reservoir site of comparable depth to the inlet site, and the third at the deepest position in the reservoir near the dam (herein referred to as the inlet, midreservoir and near-dam sites, respectively; Background, Fig. 1). Because of interest in the effects of historic hydraulic mining and ore processing in the Sierra Nevada foothills just upstream of the reservoir, dissolved-mercury species and predominant ligands that often control the mercury speciation (represented by dissolved organic carbon, and sulfides) were the solutes of primary interest. Benthic flux, sometimes referred to as internal recycling, represents the transport of dissolved chemical species between the water column and the underlying sediment. Because of the affinity of mercury to adsorb onto particle surfaces and to form insoluble precipitates (particularly with sulfides), the mass transport of mercury in mining-affected watersheds is typically particle dominated. As these enriched particles accumulate at depositional sites such as reservoirs, benthic processes facilitate the repartitioning, transformation, and transport of mercury in dissolved, biologically reactive forms (dissolved methylmercury being the most bioavailable for trophic transfer). These are the forms of mercury examined in this study. In contrast to typical scientific manuscripts, this report is formatted in a pyramid-like structure to serve the needs of diverse groups who may be interested in reviewing or acquiring information at various levels of technical detail (Appendix 1). The report enables quick transitions between the initial

  15. Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

    International Nuclear Information System (INIS)

    Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

    1979-01-01

    Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

  16. Oxidative dissolution of spent fuel and release of nuclides from a copper/iron canister. Model developments and applications

    Energy Technology Data Exchange (ETDEWEB)

    Longcheng Liu

    2001-12-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H{sub 2}, using UO{sub 2} (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H{sub 2} on the fuel surface. It will reprecipitate as UO{sub 2} (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H{sub 2} are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed

  17. 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

  18. Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

    2015-06-01

    The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  20. Alternative fuels, fuel cycles, and reactors: are they useful. are they necessary

    International Nuclear Information System (INIS)

    Spinrad, B.I.

    1985-01-01

    This chapter discusses reactors, fuel cycles, and fuel production concepts other than those considered conventional in the nuclear community. An attempt is made to look for improvements with the aim of providing cheaper and more durable energy systems, and to contribute toward a solution of the threat of weapons material diversion and weapons proliferation problems. Topics considered include breeding, alternate breeder cycles, alternative reprocessing schemes, symbiotic reactor systems, an interim strategy, and other sources of nuclear fuel. It is determined that the reprocessing of spent fuel is an important safeguard measure in itself

  1. 78 FR 20625 - Spent Nuclear Fuel Management at the Savannah River Site

    Science.gov (United States)

    2013-04-05

    ... Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact... generated at the Oak Ridge National Laboratory and approximately 1,000 bundles of aluminum-clad SNF... processing is a chemical separations process that involves dissolving spent fuel in nitric acid and...

  2. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    International Nuclear Information System (INIS)

    1978-06-01

    This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions

  3. Precipitation of plutonium from acidic solutions using magnesium oxide

    International Nuclear Information System (INIS)

    Jones, S.A.

    1994-01-01

    Magnesium oxide will be used as a neutralizing agent for acidic plutonium-containing solutions. It is expected that as the magnesium oxide dissolves, the pH of the solution will rise, and plutonium will precipitate. The resulting solid will be tested for suitability to storage. The liquid is expected to contain plutonium levels that meet disposal limit requirements

  4. Preferential removal of Sm by evaporation from Nd-Sm mixture and its application in direct burn-up determination of spent nuclear fuel

    International Nuclear Information System (INIS)

    Sajimol, R.; Bera, S.; Nalini, S.; Sivaraman, N.; Joseph, M.; Kumar, T.

    2016-01-01

    Rate of evaporation of Sm and Nd from their mixture was studied based on their ion intensities using thermal ionization mass spectrometry. Because of the comparatively larger evaporation rate of Sm, it was found possible to get the isotopic composition of Nd (fission product monitor) free from isobaric interference of Sm isotopes. The decrease in ion intensity of Sm was studied as a function of time and filament temperature. Based on this study, an easy and time effective method for the determination of burn-up of spent nuclear fuel was examined and the results are compared with that obtained by the conventional method. Typical burn-up value obtained for a pressurized heavy water reactor fuel dissolver solution using the direct method by preferential evaporation of Sm is: 0.84 at.%, whereas the one obtained by the use of conventional method is 0.82 at.%. In both the cases, Nd was employed as the fission product monitor. (author)

  5. Alternatives for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Ramirez S, J. R.; Badillo A, V.; Palacios H, J.; Celis del Angel, L.

    2010-10-01

    The spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments in the construction of repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution? or, What is the best technology for a specific solution? Many countries have deferred the decision on selecting an option, while other works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However, currently is under process an extended power up rate to 20% of their original power and also there are plans to extend operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. So this work describes some different alternatives that have been studied in Mexico to define which will be the best alternative to follow. (Author)

  6. Capability of minor nuclide confinement in fuel reprocessing

    International Nuclear Information System (INIS)

    Fujine, Sachio; Uchiyama, Gunzo; Mineo, Hideaki; Kihara, Takehiro; Asakura, Toshihide

    1999-01-01

    Experiment with spent fuels has started with the small scale reprocessing facility in NUCEF-BECKY αγ cell. Primary purpose of the experiment is to study the capability of long-lived nuclide confinement both in the PUREX flow sheet applied to the large scale reprocessing plant and also in the PARC (Partitioning Conundrum key process) flow sheet which is our proposal as a simplified reprocessing of one cycle extraction system. Our interests in the experiment are the behaviors of minor long-lived nuclides and the behaviors of the heterogeneous substances, such as sedimentation in the dissolver, organic cruds in the extraction banks. The significance of those behaviors will be assessed from the standpoint of the process safety of reprocessing for high burn-up fuels and MOX fuels. (author)

  7. Development of a SREX flowsheet for the separation of strontium from dissolved INEEL zirconium calcine

    International Nuclear Information System (INIS)

    Law, J.D.; Wood, D.J.; Todd, T.A.

    1999-01-01

    Laboratory experimentation has indicated that the SREX process is effective for partitioning 90 Sr from acidic radioactive waste solutions located at the Idaho Nuclear Technology and Engineering Center. These laboratory results were used to develop a flowsheet for countercurrent testing of the SREX process with dissolved pilot plant calcine. Testing was performed using 24 stages of 2-cm diameter centrifugal contactors which are installed in the Remote Analytical Laboratory hot cell. Dissolved Run No.64 pilot plant calcine spiked with 85 Sr was used as feed solution for the testing. The flowsheet tested consisted of an extraction section (0.15 M 4prime,4prime(5prime)-di-(tert-butylcyclohexo)-18-crown-6 and 1.5 M TBP in Isopar-L.), a 1.0 M NaNO 3 scrub section to remove extracted K from the SREX solvent, a 0.01 M HNO 3 strip section for the removal of Sr from the SREX solvent, a 0.25 M Na2CO 3 wash section to remove degradation products from the solvent, and a 0.1 M HNO 3 rinse section. The behavior of 85 Sr, Na, K, Al, B, Ca, Cr, Fe, Ni, and Zr was evaluated. The described flowsheet successfully extracted 85 Sr from the dissolved pilot plant calcine with a removal efficiency of 99.6%. Distribution coefficients for 85 Sr ranged from 3.6 to 4.5 in the extraction section. With these distribution coefficients a removal efficiency of approximately >99.99% was expected. It was determined that the lower than expected removal efficiency can be attributed to a stage efficiency of only 60% in the extraction section. Extracted K was effectively scrubbed from the SREX solvent with the 1.0 M NaNO 3 resulting in only 6.4% of the K in the HLW strip product. Sodium was not extracted from the dissolved calcine by the SREX solvent; however, the use of a 1.0 M NaNO 3 scrub solution resulted in a Na concentration of 70 mg/L (12.3% of the feed concentration) in the HLW strip product. Al, B, Ca, Cr, Fe, Ni, and Zr were determined to be essentially inextractable

  8. Final Report: Investigation of Catalytic Pathways for Lignin Breakdown into Monomers and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gluckstein, Jeffrey A [ORNL; Hu, Michael Z. [ORNL; Kidder, Michelle [ORNL; McFarlane, Joanna [ORNL; Narula, Chaitanya Kumar [ORNL; Sturgeon, Matthew R [ORNL

    2010-12-01

    Lignin is a biopolymer that comprises up to 35% of woody biomass by dry weight. It is currently underutilized compared to cellulose and hemicellulose, the other two primary components of woody biomass. Lignin has an irregular structure of methoxylated aromatic groups linked by a suite of ether and alkyl bonds which makes it difficult to degrade selectively. However, the aromatic components of lignin also make it promising as a base material for the production of aromatic fuel additives and cyclic chemical feed stocks such as styrene, benzene, and cyclohexanol. Our laboratory research focused on three methods to selectively cleave and deoxygenate purified lignin under mild conditions: acidolysis, hydrogenation and electrocatalysis. (1) Acidolysis was undertaken in CH2Cl2 at room temperature. (2) Hydrogenation was carried out by dissolving lignin and a rhodium catalyst in 1:1 water:methoxyethanol under a 1 atm H2 environment. (3) Electrocatalysis of lignin involved reacting electrically generated hydrogen atoms at a catalytic palladium cathode with lignin dissolved in a solution of aqueous methanol. In all of the experiments, the lignin degradation products were identified and quantified by gas chromatography mass spectroscopy and flame ionization detection. Yields were low, but this may have reflected the difficulty in recovering the various fractions after conversion. The homogeneous hydrogenation of lignin showed fragmentation into monomers, while the electrocatalytic hydrogenation showed production of polyaromatic hydrocarbons and substituted benzenes. In addition to the experiments, promising pathways for the conversion of lignin were assessed. Three conversion methods were compared based on their material and energy inputs and proposed improvements using better catalyst and process technology. A variety of areas were noted as needing further experimental and theoretical effort to increase the feasibility of lignin conversion to fuels.

  9. Groundwater and solute transport modeling at Hyporheic zone of upper part Citarum River

    Science.gov (United States)

    Iskandar, Irwan; Farazi, Hendy; Fadhilah, Rahmat; Purnandi, Cipto; Notosiswoyo, Sudarto

    2017-06-01

    Groundwater and surface water interaction is an interesting topic to be studied related to the water resources and environmental studies. The study of interaction between groundwater and river water at the Upper Part Citarum River aims to know the contribution of groundwater to the river or reversely and also solute transport of dissolved ions between them. Analysis of drill logs, vertical electrical sounding at the selected sections, measurement of dissolved ions, and groundwater modeling were applied to determine the flow and solute transport phenomena at the hyporheic zone. It showed the hyporheic zone dominated by silt and clay with hydraulic conductivity range from 10-4∼10-8 m/s. The groundwater flowing into the river with very low gradient and it shows that the Citarum River is a gaining stream. The groundwater modeling shows direct seepage of groundwater into the Citarum River is only 186 l/s, very small compared to the total discharge of the river. Total dissolved ions of the groundwater ranged from 200 to 480 ppm while the river water range from 200 to 2,000 ppm. Based on solute transport modeling it indicates dissolved ions dispersion of the Citarum River into groundwater may occur in some areas such as Bojongsoang-Dayeuh Kolot and Nanjung. This situation would increase the dissolved ions in groundwater in the region due to the contribution of the Citarum River. The results of the research can be a reference for further studies related to the mechanism of transport of the pollutants in the groundwater around the Citarum River.

  10. 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...

  11. Thermogravimetric analysis of fuel film evaporation

    Institute of Scientific and Technical Information of China (English)

    HU Zongjie; LI Liguang; YU Shui

    2006-01-01

    Thermogravimetric analysis (TGA) was compared with the petrochemical distillation measurement method to better understand the characteristics of fuel film evaporation at different wall tem- peratures. The film evaporation characteristics of 90# gasoline, 93# gasoline and 0# diesel with different initial thicknesses were investigated at different environmental fluxes and heating rates. The influences of heating rate, film thickness and environmental flux on fuel film evaporation for these fuels were found. The results showed that the environmental conditions in TGA were similar to those for fuel films in the internal combustion engines, so data from TGA were suitable for the analysis of fuel film evaporation. TGA could simulate the key influencing factors for fuel film evaporation and could investigate the basic quantificational effect of heating rate and film thickness. To get a rapid and sufficient fuel film evaporation, sufficiently high wall temperature is necessary. Evaporation time decreases at a high heating rate and thin film thickness, and intense gas flow is important to promoting fuel film evaporation. Data from TGA at a heating rate of 100℃/min are fit to analyze the diesel film evaporation during cold-start and warming-up. Due to the tense molecular interactions, the evaporation sequence could not be strictly divided according to the boiling points of each component for multicomponent dissolved mixture during the quick evaporation process, and the heavier components could vaporize before reaching their boiling points. The 0# diesel film would fully evaporate when the wall temperature is beyond 250℃.

  12. A Study on the Oxidative-dissolution Leaching of Fission Product Oxides in the carbonate solution

    International Nuclear Information System (INIS)

    Lee, Eil Hee; Kim, Kwang Wook; Lim, Jae Gwan; Chung, Dong Yong; Yang, Han Beom; Joe, Kih Soo; Seo, Heui Seung; Kim, Yeon Hwa; Lee, Se Yoon

    2009-07-01

    This study was carried out to investigate the characteristics of an oxidativedissolution leaching of FP co-dissolved with U in a carbonate solution of Na 2 CO 3 - H 2 O 2 and (NH 4 ) 2 CO 3 -H 2 O 2 , respectively. Simulated FP-oxides which contained 12 components have been added to the solution to examine their oxidative dissolution characteristics. It was found that H 2 O 2 was an effective oxidant to minimize the dissolution of FP in a carbonate solution. In 0.5M Na 2 CO 3 -0.5M H 2 O 2 and 0.5M (NH 4 ) 2 CO 3 -0.5M H 2 O 2 solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10∼20 minutes) due to their high solubility in Na 2 CO 3 and (NH 4 ) 2 CO 3 solution regardless of the addition of H 2 O 2 , and independent of the concentrations of Na 2 CO 3 and H 2 O 2 . However, Mo was slowly dissolved by an oxidative dissolution with H 2 O 2 . It is found that the most important factor for the oxidative dissolution of FP is the pH of the solution and an effective oxidative dissolution is achieved at a pH between 10∼12 for Na 2 CO 3 and a pH between 9∼10 for (NH 4 ) 2 CO 3 , respectively, in order to minimize the dissolution of FP

  13. Dissolution behavior of PFBR MOX fuel in nitric acid

    International Nuclear Information System (INIS)

    Kelkar, Anoop; Kapoor, Y.S.; Singh, Mamta; Meena, D.L.; Pandey, Ashish; Bhatt, R.B.; Behere, P.G.

    2017-01-01

    Present paper describes the dissolution characteristics of PFBR MOX fuel (U,Pu)O 2 in nitric acid. An overview of batch dissolution experiments, studying the percentage dissolution of uranium and plutonium in (U, Pu)O 2 MOX sintered pellets with different percentage of PuO 2 with reference to time and nitric acid concentration are described. 90% of uranium and plutonium of PFBR MOX gets dissolves in 2 hrs and amount of residue increases with the decrease in nitric acid concentration. Overall variation in percentage residue in PFBR MOX fuel after dissolution test also described. (author)

  14. Reaction of hydrogen peroxide with uranium zirconium oxide solid solution - Zirconium hinders oxidative uranium dissolution

    Science.gov (United States)

    Kumagai, Yuta; Takano, Masahide; Watanabe, Masayuki

    2017-12-01

    We studied oxidative dissolution of uranium and zirconium oxide [(U,Zr)O2] in aqueous H2O2 solution to estimate (U,Zr)O2 stability to interfacial reactions with H2O2. Studies on the interfacial reactions are essential for anticipating how a (U,Zr)O2-based molten fuel may chemically degrade after a severe accident. The fuel's high radioactivity induces water radiolysis and continuous H2O2 generation. Subsequent reaction of the fuel with H2O2 may oxidize the fuel surface and facilitate U dissolution. We conducted our experiments with (U,Zr)O2 powder (comprising Zr:U mole ratios of 25:75, 40:60, and 50:50) and quantitated the H2O2 reaction via dissolved U and H2O2 concentrations. Although (U,Zr)O2 reacted more quickly than UO2, the dissolution yield relative to H2O2 consumption was far less for (U,Zr)O2 compared to that of UO2. The reaction kinetics indicates that most of the H2O2 catalytically decomposed to O2 at the surface of (U,Zr)O2. We confirmed the H2O2 catalytic decomposition via O2 production (quantitative stoichiometric agreement). In addition, post-reaction Raman scattering spectra of the undissolved (U,Zr)O2 showed no additional peaks (indicating a lack of secondary phase formation). The (U,Zr)O2 matrix is much more stable than UO2 against H2O2-induced oxidative dissolution. Our findings will improve understanding on the molten fuels and provide an insight into decommissioning activities after a severe accident.

  15. 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.

  16. A Monte Carlo simulation method for assessing biotransformation effects on groundwater fuel hydrocarbon plume lengths

    International Nuclear Information System (INIS)

    McNab, W.W. Jr.

    2000-01-01

    Biotransformation of dissolved groundwater hydrocarbon plumes emanating from leaking underground fuel tanks should, in principle, result in plume length stabilization over relatively short distances, thus diminishing the environmental risk. However, because the behavior of hydrocarbon plumes is usually poorly constrained at most leaking underground fuel tank sites in terms of release history, groundwater velocity, dispersion, as well as the biotransformation rate, demonstrating such a limitation in plume length is problematic. Biotransformation signatures in the aquifer geochemistry, most notably elevated bicarbonate, may offer a means of constraining the relationship between plume length and the mean biotransformation rate. In this study, modeled plume lengths and spatial bicarbonate differences among a population of synthetic hydrocarbon plumes, generated through Monte Carlo simulation of an analytical solute transport model, are compared to field observations from six underground storage tank (UST) sites at military bases in California. Simulation results indicate that the relationship between plume length and the distribution of bicarbonate is best explained by biotransformation rates that are consistent with ranges commonly reported in the literature. This finding suggests that bicarbonate can indeed provide an independent means for evaluating limitations in hydrocarbon plume length resulting from biotransformation. (Author)

  17. The cost of spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Palacios H, J. C.; Badillo, V.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    Spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments, constructing repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution?, or What is the best technology for an specific solution? Many countries have deferred the decision on selecting an option, while others works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However currently, the plants are under a process for extended power up-rate to 20% of original power and also there are plans to extended operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. (Author)

  18. Cold Dissolved Saltcake Waste Simulant Development, Preparation, and Analysis

    International Nuclear Information System (INIS)

    Rassat, Scot D.; Mahoney, Lenna A.; Russell, Renee L.; Bryan, Samuel A.; Sell, Rachel L.

    2003-01-01

    CH2M HILL Hanford Group, Inc. is identifying and developing supplemental process technologies to accelerate the Hanford tank waste cleanup mission. Bulk vitrification, containerized grout, and steam reforming are three technologies under consideration for treatment of the radioactive saltcake wastes in 68 single-shell tanks. To support development and testing of these technologies, Pacific Northwest National Laboratory (PNNL) was tasked with developing a cold dissolved saltcake simulant formulation to be representative of an actual saltcake waste stream, preparing 25- and 100-L batches of the simulant, and analyzing the composition of the batches to ensure conformance to formulation targets. Lacking a defined composition for dissolved actual saltcake waste, PNNL used available tank waste composition information and an equilibrium chemistry model (Environmental Simulation Program [ESP(trademark)]) to predict the concentrations of analytes in solution. Observations of insoluble solids in initial laboratory preparations for the model-predicted formulation prompted reductions in the concentration of phosphate and silicon in the final simulant formulation. The analytical results for the 25- and 100-L simulant batches, prepared by an outside vendor to PNNL specifications, agree within the expected measurement accuracy (∼10%) of the target concentrations and are highly consistent for replicate measurements, with a few minor exceptions. In parallel with the production of the 2nd simulant batch (100-L), a 1-L laboratory control sample of the same formulation was carefully prepared at PNNL to serve as an analytical standard. The instrumental analyses indicate that the vendor prepared batches of solution adequately reflect the as-formulated simulant composition. In parallel with the simulant development effort, a nominal 5-M (molar) sodium actual waste solution was prepared at the Hanford Site from a limited number of tank waste samples. Because this actual waste solution w

  19. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    Myllykylae, E.

    2008-02-01

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  20. A new coupled system for BWR nuclear fuel management

    International Nuclear Information System (INIS)

    Castillo, A.; Ortiz-Servin, J.J.; Montes-Tadeo, J.L.; Perusquia, R.; Rizos, R.L.M.

    2015-01-01

    In this work, a system to solve four stages of the fuel management problem is showed.The system uses different heuristic techniques to solve each stage of that area, and this problem is solved in a coupled way. Considered problems correspond to the following designs: fuel lattice, fuel assembly, fuel reload and control rod patterns. Even though, each stage of the problem can have its own objective function, the complete problem was solved using a multi-objective function. The solution strategy is to solve each stage of design in an iterative process, taking into account previous results for the next stage, until to achieve a complete solution. The solution strategy to solve the coupled problem is the following: the first solved stage is the fuel lattice design, the second one is fuel assembly design, finally an internal loop between both fuel reload design and control rod pattern design is carried out.For this internal loop, a seed reload using Haling principle is generated. The obtained results showed the advantage to solve the whole problem in a coupled way. (author)

  1. 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.)

  2. Use of a dissolved-gas measurement system for reducing the dissolved oxygen at St. Lucie Unit 2

    International Nuclear Information System (INIS)

    Snyder, D.T.; Coit, R.L.

    1993-02-01

    When the dissolved oxygen in the condensate at St. Lucie Unit 2 could not be reduced below the administrative limit of 10 ppB, EPRI cooperated with Florida Power and Light to find the cause and develop remedies. Two problems were identified with the assistance of a dissolved gas measurement system (DGMS) that can detect leaks into condensate when used with argon blanketing. Drain piping from the air ejection system had flooded which decreased its performance, and leaks were found at a strainer flange and a couple expansion joints. Initially the dissolved oxygen content was reduced to about 9 ppB; owever, the dissolved oxygen from Condenser A was consistently higher than that from condenser B. Injection of about 0.4 cubic per minute (CFM) of argon above the hotwell considerably improved the ventilation of Condenser A, reducing the dissolved oxygen about 30% to about 6 ppB. The use of nitrogen was equally effective. While inert gas injection is helpful, it may be better to have separate air ejectors for each condenser. Several recommendations for improving oxygen removal are given

  3. Process for recovering uranium using an alkyl pyrophosphoric acid and alkaline stripping solution

    International Nuclear Information System (INIS)

    Worthington, R.E.; Magdics, A.

    1987-01-01

    A process is described for stripping uranium for a pregnant organic extractant comprising an alkyl pyrophosphoric acid dissolved in a substantially water-immiscible organic diluent. The organic extractant contains tetravalent uranium and an alcohol or phenol modifier in a quantity sufficient to retain substantially all the unhydrolyzed alkyl pyrophosphoric acid in solution in the diluent during stripping. The process comprises adding an oxidizing agent to the organic extractant and thereby oxidizing the tetravalent uranium to the +6 state in the organic extractant, and contacting the organic extractant containing the uranium in the +6 state with a stripping solution comprising an aqueous solution of an alkali metal or ammonium carbonate or hydroxide thereby stripping uranium from the organic extractant into the stripping solution. The resulting barren organic extractant containing substantially all of the unhydrolyzed alkyl pyrophosphoric acid dissolved in the diluent is separated from the stripping solution containing the stripped uranium, the barren extractant being suitable for recycle

  4. Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System

    International Nuclear Information System (INIS)

    Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara; Kazunori Nomura; Shigehiko Miyachi; Yoshiaki Ichige; Tadahiro Shinozaki; Shinichi Ohuchi

    2008-01-01

    CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation

  5. Road transport fuels in europe: the explosion of demand for diesel fuel

    International Nuclear Information System (INIS)

    Bensaid, B.

    2004-01-01

    In the last 20 years, road transport fuel consumption has more than doubled in European countries, due to strong growth on the diesel passenger car segment and in the transport of road freight. In an economy heavily dependent on oil, European authorities are seeking to promote alternative energy solutions, such as motor fuels produced from biomass

  6. Membrane separation of ionic liquid solutions

    Science.gov (United States)

    Campos, Daniel; Feiring, Andrew Edward; Majumdar, Sudipto; Nemser, Stuart

    2015-09-01

    A membrane separation process using a highly fluorinated polymer membrane that selectively permeates water of an aqueous ionic liquid solution to provide dry ionic liquid. Preferably the polymer is a polymer that includes polymerized perfluoro-2,2-dimethyl-1,3-dioxole (PDD). The process is also capable of removing small molecular compounds such as organic solvents that can be present in the solution. This membrane separation process is suitable for drying the aqueous ionic liquid byproduct from precipitating solutions of biomass dissolved in ionic liquid, and is thus instrumental to providing usable lignocellulosic products for energy consumption and other industrial uses in an environmentally benign manner.

  7. 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

  8. 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

  9. 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

  10. Improvement of shacking helical elevators used in spent fuel reprocessing. Perfectionnement aux elevateurs helicoidaux a secousses, utilises dans le traitement des combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

    Tucoulat, D.; Kerlau, D.; Cagin, R.; Pellier, R.; Tarnero, M.; Saudray, D.

    1991-01-25

    For reprocessing cut spent fuel elements are introduced in a tank and raised gradually with an helical ramp by a back and forth motion around a vertical axis. Spent fuel is dissolved and hulls are recovered at the top of the ramp.

  11. Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

    Directory of Open Access Journals (Sweden)

    A. V. Zolin

    2015-01-01

    Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

  12. Device for isolation of seed crystals during processing of solution

    Science.gov (United States)

    Montgomery, K.E.; Zaitseva, N.P.; Deyoreo, J.J.; Vital, R.L.

    1999-05-18

    A device is described for isolation of seed crystals during processing of solutions. The device enables a seed crystal to be introduced into the solution without exposing the solution to contaminants or to sources of drying and cooling. The device constitutes a seed protector which allows the seed to be present in the growth solution during filtration and overheating operations while at the same time preventing the seed from being dissolved by the under saturated solution. When the solution processing has been completed and the solution cooled to near the saturation point, the seed protector is opened, exposing the seed to the solution and allowing growth to begin. 3 figs.

  13. Corrosion Behavior and Oxide Properties of Zr-Nb-Cu and Zr-Nb-Sn Alloy in High Dissolved Hydrogen Primary Water Chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yun Ju; Kim, Tae Ho; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    The water-metal interface is regarded as rate-controlling site governing the rapid oxidation transition in high burn-up fuel. And the zirconium oxide is made in water-metal interface and its structure and phase do an important role in terms of oxide properties. During oxidation process, the protective tetragonal oxide layer develops at the interface due to accumulated high stress during oxide growth, and it turns into non-protective monoclinic oxide with increasing oxide thickness, thus decreasing the stress. It has been reported that Nb addition was proven to be very beneficial for increasing the corrosion resistance of the zirconium alloys. From a more recent study, Cu addition in Nb containing Zirconium alloy was reported to be effective for increasing corrosion resistance in water containing B and Li. According to the previous research conducted, Zr-Nb-Cu shows better corrosion resistance than Zircaloy-4. The dissolved hydrogen (DH) concentration is the key issue of primary water chemistry, and the effect of DH concentration on the corrosion rate of nickel based alloy has been researched. However, the effect of DH on the zirconium alloy corrosion mechanism was not fully investigated. In this study, the weight gain measurement, FIB-SEM analysis, and Raman spectroscopic measurement were conducted to investigate the effects of dissolved hydrogen concentration and the chemical composition on the corrosion resistance and oxide phase of Zr-Nb-Cu alloy and Zr-Nb-Sn alloy after oxidizing in a primary water environment for 20 d. The corrosion rate of Zr-Nb-Cu alloy is slow, when it is compared to Zr-Nb-Sn alloy. In SEM images, the oxide thickness of Zr-Nb-Cu alloy is measured to be around 1.06 μm it of Zr-Nb-Sn alloy is measured to be 1.15 μm. It is because of the Segregation made by Sn solute element when Sn solute element oxidized. And according to ex situ Raman spectra, Zr-Nb-Cu alloy oxide has more tetragonal zirconium oxide fraction than Zr-Nb-Sn alloy oxide.

  14. 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)

  15. 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.

  16. Enhanced Indirect Photochemical Transformation of Histidine and Histamine through Association with Chromophoric Dissolved Organic Matter.

    Science.gov (United States)

    Chu, Chiheng; Lundeen, Rachel A; Remucal, Christina K; Sander, Michael; McNeill, Kristopher

    2015-05-05

    Photochemical transformations greatly affect the stability and fate of amino acids (AAs) in sunlit aquatic ecosystems. Whereas the direct phototransformation of dissolved AAs is well investigated, their indirect photolysis in the presence of chromophoric dissolved organic matter (CDOM) is poorly understood. In aquatic systems, CDOM may act both as sorbent for AAs and as photosensitizer, creating microenvironments with high concentrations of photochemically produced reactive intermediates, such as singlet oxygen (1O2). This study provides a systematic investigation of the indirect photochemical transformation of histidine (His) and histamine by 1O2 in solutions containing CDOM as a function of solution pH. Both His and histamine showed pH-dependent enhanced phototransformation in the CDOM systems as compared to systems in which model, low-molecular-weight 1O2 sensitizers were used. Enhanced reactivity resulted from sorption of His and histamine to CDOM and thus exposure to elevated 1O2 concentrations in the CDOM microenvironment. The extent of reactivity enhancement depended on solution pH via its effects on the protonation state of His, histamine, and CDOM. Sorption-enhanced reactivity was independently supported by depressed rate enhancements in the presence of a cosorbate that competitively displaced His and histamine from CDOM. Incorporating sorption and photochemical transformation processes into a reaction rate prediction model improved the description of the abiotic photochemical transformation rates of His in the presence of CDOM.

  17. R and D on fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Subba Rao, R.V.; Vijaya Kumar, V.; Natarajan, R.

    2012-01-01

    Development of Fast Reactor Fuel Reprocessing technology, with low out of pile inventory, is carried out at the Indira Gandhi Centre for Atomic Research (IGCAR). Based on the successful R and D programme which addressed specific issues of fast reactor fuels, a pilot plant called CORAL was set up. This plant is operational since 2003 and several reprocessing campaigns with spent FBTR fuels of varying burnups have been carried out. Based on the valuable operating experience of CORAL, the design of demonstration fast reactor fuel reprocessing plant (DFRP) and the commercial reprocessing plant, FRP have been taken up. Concurrently R and D efforts are continuing for improving the process and equipment performance apart from reducing the waste volumes and the radiation exposures to the operating personnel. Some important R and D efforts are highlighted in the paper. Reducing the dissolution time is one of the vital area of investigation especially for the high plutonium bearing MOX fuels which are known to dissolve slowly. To address this as well as criticality issues, continuous dissolvers are being developed. Solvent extraction based process is employed for getting highly pure nuclear grade uranium and plutonium. In view of the lower cooling time the fission product activity in the spent fuel is higher, formulation of process flowsheet with reduced number of solvent extraction cycles to improve the decontamination of ruthenium and zirconium without the formation of second organic phase due to plutonium loading, is under investigation. Retention of plutonium in lean organic is another issue to be addressed as otherwise it would lead to further deterioration of the solvent on storage. Several reagents to effectively wash the lean solvent have been investigated and flowsheets have been formulated to recover the retained plutonium with minimum secondary wastes. Partitioning of uranium and plutonium is an important step and methods reported in the literature have inherent

  18. 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.

  19. 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.

  20. Radiation-chemical degradation of chloroform in water solutions

    International Nuclear Information System (INIS)

    Ahmadov, S.A.; Gurbanov, M.A.; Iskenderova, Z.I.; Abdullayev, E.T.; Ibadov, N.A.

    2006-01-01

    Full text: Chloroform is the major chlorine-containing compound forming at chlorination of drinking water. As our basic water resources of Kur and Araz rivers are mostly polluted along the territory of the neighbour republics their chlorination for the purpose of biological purification can result in forming of chloroform. Unfortunately, there are only poor data about containing of chloroform in drinking water in the Republic, however the particular problem is to develop new methods of drinking water purification from chloroform, taking into account the high toxicity of this compounds. Appropriate works indicate that radiation-chemical processing can mostly reduce the concentration of chloroform in drinking water. The purification degree can achieve 95-98 percent. This work studies the tendency of chloroform decomposition at its radiolysis processes in water solutions. The concentration of chloroform changed in the range of 0,03-1 weight percentage. Taking into account the dissolvability of chloroform in water solutions it can be said that examined water solutions are homogeneous. Following advancements are studied: 1) Determination of radiation-chemical yield of chloroform decomposition at its various initial concentrations; 2) Impact of adsorbed dose on pH of solutions; 3) Formation of by-products. It is set that radiation-chemical output of chloroform decomposition is equal to 3 * 10 - 3 - 125 mol/100 ev. The high yield of chloroform decomposition can be connected with the chain process of oxidation with presence of dissolved oxygen. However, taking into account the fact that at its water radiolysis the yield of active particles of OH, e - aq, H-atoms does not exceed 6-7 particles/100 ev, the observed high yield can be explained only with the chain process with presence of dissolved oxygen