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Sample records for fissium

  1. Development of a reliable analytical method for extraction spectrophotometric determination of ruthenium(III) from catalyst and fissium alloy using o-methylphenyl thiourea as a chromogenic chelating ligand

    NASA Astrophysics Data System (ADS)

    Kuchekar, Shashikant R.; Shelar, Yogesh S.; Aher, Haribhau R.; Han, Sung H.

    2013-04-01

    A simple and selective method is developed for the extraction spectrophotometric determination of ruthenium(III) using o-methylphenyl thiourea (OMPT) as a chromogenic chelating ligand. The basis of the proposed method is ruthenium(III)-OMPT complex formation in aqueous hydrochloric acid media (3.0 mol L-1) after 5.0 min heating on a boiling water bath and the complex formed is extracted into chloroform. The absorbance of green colored ruthenium(III)-OMPT complex is measured at 590 nm against the reagent blank. Beer's law was obeyed up to 42.5 μg mL-1 of ruthenium(III) and the optimum concentration range is 7.56-39.81 μg mL-1 of ruthenium(III) as evaluated by Ringbom's plot. Molar absorptivity and Sandell's sensitivity of ruthenium(III)-OMPT complex in chloroform are 2.34 × 103 L mol-1 cm-1 and 0.043 μg cm-2 respectively. The composition of ruthenium(III):OMPT complex (1:2) was established from slope ratio method, mole ratio method and Job's continuous variation method. Complex was stable for more than 48 h. The interfering effect of various foreign ions was studied and suitable masking agents are used wherever necessary to enhance the selectivity of the method. Proposed method is successfully applied for determination of ruthenium(III) from binary and ternary synthetic mixtures, synthetic mixtures corresponding to fissium alloy and ruthenium catalyst. Repetition of the method was checked by finding relative standard deviation (R.S.D) for 10 determinations which was 0.23%. A scheme for sequential separation of palladium(II), ruthenium(III), rhodium(III) and platinum(IV) has been developed.

  2. Development of a reliable analytical method for extraction spectrophotometric determination of ruthenium(III) from catalyst and fissium alloy using o-methylphenyl thiourea as a chromogenic chelating ligand.

    PubMed

    Kuchekar, Shashikant R; Shelar, Yogesh S; Aher, Haribhau R; Han, Sung H

    2013-04-01

    A simple and selective method is developed for the extraction spectrophotometric determination of ruthenium(III) using o-methylphenyl thiourea (OMPT) as a chromogenic chelating ligand. The basis of the proposed method is ruthenium(III)-OMPT complex formation in aqueous hydrochloric acid media (3.0molL(-1)) after 5.0min heating on a boiling water bath and the complex formed is extracted into chloroform. The absorbance of green colored ruthenium(III)-OMPT complex is measured at 590nm against the reagent blank. Beer's law was obeyed up to 42.5μgmL(-1) of ruthenium(III) and the optimum concentration range is 7.56-39.81μgmL(-1) of ruthenium(III) as evaluated by Ringbom's plot. Molar absorptivity and Sandell's sensitivity of ruthenium(III)-OMPT complex in chloroform are 2.34×10(3)Lmol(-1) cm(-1) and 0.043μgcm(-2) respectively. The composition of ruthenium(III):OMPT complex (1:2) was established from slope ratio method, mole ratio method and Job's continuous variation method. Complex was stable for more than 48h. The interfering effect of various foreign ions was studied and suitable masking agents are used wherever necessary to enhance the selectivity of the method. Proposed method is successfully applied for determination of ruthenium(III) from binary and ternary synthetic mixtures, synthetic mixtures corresponding to fissium alloy and ruthenium catalyst. Repetition of the method was checked by finding relative standard deviation (R.S.D) for 10 determinations which was 0.23%. A scheme for sequential separation of palladium(II), ruthenium(III), rhodium(III) and platinum(IV) has been developed. Copyright © 2013 Elsevier B.V. All rights reserved.

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

    NASA Astrophysics Data System (ADS)

    Feuerstein, H.; Oschinski, J.

    1986-11-01

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

  4. DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

    DOEpatents

    Zegler, S.T.

    1963-11-01

    The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

  5. Recent metal fuel safety tests in TREAT

    SciTech Connect

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described.

  6. Local fault tolerance of metal fuel

    SciTech Connect

    Tilbrook, R W; Pedersen, D R; Thompson, D H; Ragland, W A

    1986-02-01

    This IFR technical memorandum presents a review of the potential initiators of fuel failure in metal fuel and a preliminary evaluation of the consequences of failure and the potential for propagation within a fuel assembly. Lines of defense against initiation and propagation are identified and some discussed in detail including appropriate supportive conclusions from oxide fuel assessments. The ongoing supporting fuel element test program is described and areas requiring further analytical or experimental effort are identified. Based on the extensive experience in EBR-II with uranium-fissium fuel, and the differences between the properties of metallic and oxide fuel constitutents, superior local faults tolerance of ternary alloy fuel is anticipated. 34 refs.

  7. Posttest examination results of recent treat tests on metal fuel

    SciTech Connect

    Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

    1986-01-01

    A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins.

  8. Integral fast reactor safety tests M2 and M3 in TREAT

    SciTech Connect

    Robinson, W.R.; Lo, R.K.; Wright, A.E.; Bauer, T.H.; Stanford, G.S.; Morman, J.A.

    1985-11-01

    Transient Reactor Test Facility (TREAT) tests M2 and M3 were performed to obtain information on two key fuel behavior characteristics of transient overpower conditions in metal-fueled fast reactors: the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Cladding breach depends on penetration of a fuel/cladding eutectic into the cladding as well as on internal pin pressure. Driving forces for fuel extrusion are fission gas, liquid sodium, and volatile fission products trapped within the fuel. Significant fuel extrusion prior to cladding breach would be an important factor in the case for benign termination of unprotected overpower events in a fast reactor. These preliminary tests in the Integral Fast Reactor (IFR) program were done on uranium-5% fissium Experimental Breeder Reactor II Mark-II driver fuel pins having an active fuel column length of 34 cm.

  9. Cladding failure margins for metallic fuel in the integral fast reactor

    SciTech Connect

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel.

  10. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  11. Review of fuel/cladding eutectic formation in metallic SFR fuel pins

    SciTech Connect

    Denman, M.; Todreas, N.; Driscoll, M.

    2012-07-01

    Sodium-cooled Fast Reactors (SFRs) remain a strong contender amongst the Generation IV reactor concepts. Metallic fuel has been a primary fuel option for SFR designers in the US and was used extensively in the first generation of SFRs. One of the benefits of metallic fuel is its chemical compatibility with the coolant; unfortunately this compatibility does not extend to steel cladding at elevated temperatures. It has been known that uranium, plutonium, and rare earths diffuse with cladding constituents to form a low melting point fuel/cladding eutectic which acts to thin the cladding once the interfacial temperature rises above the system liquidus temperature. Since the 1960's, many experiments have been performed and published to evaluate the rate of fuel/cladding eutectic formation and the temperature above which melting will begin as a function of fuel/cladding interfacial temperature, time at temperature, fuel constituents (uranium, fissium or uranium (plutonium) zirconium), cladding type (stainless steel 316, stainless steel 306, D9 or HT9), beginning of life linear power, plutonium enrichment and burnup. The results of these tests, however, remain scattered across conference and journal papers spanning 50 years. The tests used to collect this data also varied in experimental procedure throughout the years. This paper will consolidate the experimental data into four groups of similar test conditions and expand upon the testing performed for each group in detail. A companion paper in PSA 2011 will discuss predictive correlations formulated from this database. (authors)

  12. EBR-II metallic driver fuel - a live option

    SciTech Connect

    Seidel, B.R.; Walters, L.C.

    1981-10-01

    The exceptional performance of metallic driver fuel has been demonstrated by the irradiation of a large number of Experimental Breeder Reactor II (EBR-II) driver-fuel elements of uranium-5 wt percent fissium clad in austenitic stainless steel. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. The irradiation performance has been improved by decreasing the fuel smear density, increasing the plenum volume, increasing the cladding thickness, and selecting a higher-strength, lower-swelling cladding alloy which exhibits less fuel-cladding chemical interaction. Quantification of reliability has allowed full utilization of the element lifetime. Lifetimes much greater than 10 at. percent could be achieved by a design change of the restrainer, which currently limits life. Use of U-Pu-Zr fuel alloy with current cladding material would provide higher-temperature capability. Metallic fuel systems with their inherently superior breeding and irradiation performance are a capable and attractive next-generation power systems. 19 refs.