Development of `DUCRETE`

This interim report covers theoretical and experimental aspects of a series of scoping tests using depleted uranium oxide pieces as aggregate in portland cement to form concrete (DUCRETE). DUCRETE is expected to provide a high integrity material suitable for shielding in spent nuclear fuel containers or for direct disposal in a low- level waste repository. The uranium oxide would produced by conversion of depleted UF{sub 6} stored by the Department of Energy

Development of DU-AGG (Depleted Uranium Aggregate)

Depleted uranium oxide (UO{sub 2} or U0{sub 3}) powder was mixed with fine mineral additives, pressed, and heated to about 1,250{degree}C. The additives were chemically constituted to result in an iron-enriched basalt (IEB). Melting and wetting of the IEB phase caused the urania powder compact to densify (sinter) via a liquid phase sintering mechanism. An inorganic lubricant was found to aid in green-forming of the body. Sintering was successful in oxidizing (air), inert (argon), or reducing (dry hydrogen containing) atmospheres. The use of ground U0{sub 3} powders (93 vol %) followed by sintering in a dry hydrogen-containing atmosphere significantly increased the density of samples (bulk density of 8.40 g/cm{sup 3} and apparent density of 9.48 g/cm{sup 3}, open porosity of 11.43%). An improvement in the microstructure (reduction in open porosity) was achieved when the vol % of U0{sub 3} was decreased to 80%. The bulk density increased to 8.59 g/cm{sup 3}, the apparent density decreased slightly to 8.82 g/cm{sup 3} (due to increase of low density IEB content), while the open porosity decreased to an excellent number of 2.78%. A representative sample derived from 80 vol % U0{sub 3} showed that most pores were closed pores and that, overall, the sample achieved the excellent relative density value of 94.1% of the estimated theoretical density (composite of U0{sub 2} and IEB). It is expected that ground powders of U0{sub 3} could be successfully used to mass produce lowcost aggregate using the green-forming technique of briquetting

DU-AGG pilot plant design study

The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule

Structural Ceramic Composites for Nuclear Applications

A research program has been established to investigate fiber reinforced ceramic composites to be used as control rod components within a Very High Temperature Reactor. Two candidate systems have been identified, carbon fiber reinforced carbon (Cf/C) and silicon carbide fiber reinforced silicon carbide (SiCf/SiC) composites. Initial irradiation stability studies to determine the maximum dose for each composite type have been initiated within the High Flux Isotope Reactor at Oak Ridge National Laboratory. Test samples exposed to 10 dpa irradiation dose have been completed with future samples to dose levels of 20 and 30 dpa scheduled for completion in following years. Mechanical and environmental testing is being conducted concurrently at the Idaho National Laboratory and at Pacific Northwest National Laboratory. High temperature test equipment, testing methodologies, and test samples for high temperature (up to 1600º C) tensile strength and long duration creep studies have been established. Specific attention was paid to the architectural fiber preform design as well as the materials used in construction of the composites. Actual testing of both tubular and flat, "dog-bone" shaped tensile composite specimens will begin next year. Since there is no precedence for using ceramic composites within a nuclear reactor, ASTM standard test procedures will be established from these mechanical and environmental tests. Close collaborations between the U.S. national laboratories and international collaborators (i.e. France and Japan) are being forged to establish both national and international test standards to be used to qualify ceramic composites for nuclear reactor applications

High temperature fuel cell research and development. Final technical status report, 1 March 1979-31 March 1980. [Molten carbonate fuel cells]

The objective of the program has been to evaluate potential materials for their chemical and physical stability in molten carbonate fuel cell electrolyte. Eleven powdered candidate materials were selected based on previous physical and chemical stability tests at elevated temperatures on solid materials and/or their thermodynamic stability with respect to proposed degradation reactions. The eleven candidate materials, plus gamma lithium aluminate, were characterized prior to corrosion testing utilizing (a) Chemical Analysis, (b) X-ray Diffraction for Phase Identification, (c) Scanning Electron Microscopy (SEM), and (d) Surface Area Analysis (B.E.T.). The powders were corrosion tested initially for 200 hours by heating to 700/sup 0/C in a mixture of 62% Li/sub 2/CO/sub 3/-38% K/sub 2/CO/sub 3/ under a fuel gas atmosphere. The gas composition was based on reformed Naphtha at 700/sup 0/C and consisted of 50.2% H/sub 2/, 10.8% CO, 9.5% CO/sub 2/, and 29.5% H/sub 2/O. The samples were tested in an inert container made by coating the interior of alumina crucibles with a layer of 0.002-inch gold. Seven of the original test materials (Al rich MgAl/sub 2/O/sub 4/, Mg rich MgAl/sub 2/O/sub 4/, Al/sub 2/(/sub 1-x/)Cr/sub 2x/O/sub 3/, CaZrO/sub 3/, MgTa/sub 2/O/sub 6/, CoAl/sub 2/O/sub 4/, and MgZrO/sub 3/) were eliminated from further testing after completion of the 200-hour test. The remaining materials (ZrO/sub 2/ stabilized with CaO, ZrO/sub 2/ stabilized with Y/sub 2/O/sub 3/ and Cr/sub 2/O/sub 3/, Y/sub 2/O/sub 3/ . Al/sub 2/O/sub 3/, ..gamma..-LiAlO/sub 2/, ThO/sub 2/, and CeO/sub 2/ as a potential substitute for ThO/sub 2/) were corrosion tested for 1000 hours under the same conditions as the previous 200-hour test. Results are presented and discussed. Powders of ZrO/sub 2/ stabilized with Y/sub 2/O/sub 3/ and Cr/sub 2/O/sub 3/, ThO/sub 2/, and ..gamma..-LiAlO/sub 2/ show the most corrosion resistance to 62% Li/sub 2/CO/sub 3/-38% K/sub 2/CO/sub 3/ at 700/sup 0/C under a reformed Naphtha fuel gas atmosphere for 1000 hours. (WHK

Structural characterization and ferroelectric properties of strontium barium niobate (Sr xBa1-xNb2O6) thin films

Strontium barium niobate (SBN) thin films of good quality were deposited on Pt/Ti/SiO2/Si substrate using a polymeric resin containing metallic ions. Films were crystallized at different temperatures and for different duration of time. The structure of these films was studied using X-ray diffraction. The coexistence of SrNb2O6 (SN) and SBN was observed in films crystallized at 700 °C. The amount of SN decreases when the crystallization time increases. Ferroelectric properties were determined for films crystallized at 700 °C for 1 and 5 h. For SBN film crystallized at 700 °C for 1 h, the remanent polarization (Pr) and the coercive field (Ec) were 2.6 muC/cm² and 71.9 kV/cm, respectively. For the film crystallized at 700 °C for 5 h these parameters were Pr = 1.1 muC/cm² and Ec = 50.5 kV/cm

Ferroelectric thin films using oxides as raw materials

This work describes an alternative method for the preparation of ferroelectric thin films based on pre-calcination of oxides, to be used as precursor material for a solution preparation. In order to show the viability of the proposed method, PbZr0.53Ti0.47O3 and Bi4Ti3O12 thin films were prepared on fused quartz and Si substrates. The results were analyzed by X-ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Infrared Spectroscopy (IR) and Rutherford Backscattering Spectroscopy (RBS). The films obtained show good quality, homogeneity and the desired stoichiometry. The estimated thickness for one layer deposition was approximately 1000 Å and 1500 Å for Bi4Ti3O12 and PbZr0.53Ti0.47O3 films, respectively