[en] Hydrous metal oxides of Zr, Ti, Hf, Fe, Al, etc. are inorganic ion exchangers that have high selectivities and efficiencies for separating and removing fission products, actinides, and other undesirable elements from aqueous waste streams. In most cases, these ion exchangers are commercially available only as fine powders or as unstable granular particles that are not readily adaptable to continuous processing techniques such as column chromatography. Hydrous metal oxides can be prepared as microspheres by the internal gelation process. This process is unique in that it provides a means of making a usable engineered form of inorganic ion exchanger that can be used in large-scale column separations. With such material, the flow dynamics in column operations would be greatly enhanced. In addition, the microspheres are in a stable form that has little or no tendency to degrade under dynamic conditions. Another advantage of the process is that the gelation time and size of the microspheres can be controlled. Also, microspheres can be reproducibly prepared on either a small or a large scale-which is not always true for batch preparation of the powdered or granular forms. The use of these materials can be expanded in a number of ways. The process allows for the microspheres to be homogeneously embedded with fine particles of other selective ion exchangers, and for the microspheres (undried) to be chemically converted to microspheres of other ion-exchanger materials such as phosphates, silicophosphates, hexacyanoferrates, tungstates, and molybdates. This report presents an economic evaluation of the preparation of hydrous titanium oxide (HTiO) microspheres by an internal gelation process for use in making ion exchangers, catalysts, and getters. It also examines the estimated costs for a company to produce the material but does not discuss the price to be charged since that value would take into account company policy-matters that cannot be covered here. Since the volume of business is not known, the costs were bracketed between the laboratory-scale system of making 1 to 2 lb HTiO/d of dried beads per 8-h day and a small pilot-scale system of producing 1 to 2.4 lb HTiO/h. The best estimates were between $286 and $534 for the laboratory-scale production of 520 and 260 lb/year, respectively, and between $93 and $107 for the pilot-scale production of 1.5 tons/year. The costs of producing microspheres in a pilot-scale facility will be strongly dependent on the scale of the facility and the fraction of time it is used. The preparation of inorganic materials as microspheres has the potential for many additional applications. If these applications prove to be feasible, the cost of producing the materials could be decreased even further

[en] In the development of (U,Pu)O₂ kernels by the internal gelation process for the Direct Press Spheroidized process at Oak Ridge National Laboratory, a novel crystal growth step was discovered that made it possible to prepare calcined porous kernels that could be used as direct-press feed for Fast Breeder Reactor pellet fabrication. High-quality pellets were prepared that were near theoretical density and that (upon examination) revealed no evidence of sphere remnants. The controlled crystal growth step involved using hexamethylenetetramine (HMTA)-urea stock solutions that were boiled for 60 min or less. Before this discovery, all the other crystal growth steps (when utilized) could reduce the tap density to only ∼1.3 g/cm³, which was not sufficiently low for use in ideal pellet pressing. The use of the boiled HMTA-urea solution allowed the tap density to be lowered to 0.93 g/cm³, with the ideal density being about 1.0 g/cm³. This report describes the development of this technology and its scaleup

[en] The effects of various radiations on commercially made Al-SiO₂-Si Capacitors (MOSCs) have been investigated. Intrinsic dielectric breakdown in MOSCs has been shown to be a two-stage process dominated by charge injection in a pre-breakdown stage; this is associated with localised high-field injection of carriers from the semiconductor substrate to interfacial and bulk charge traps which, it is proposed, leads to the formation of conducting channels through the dielectric with breakdown occurring as a result of the dissipation of the conduction band energy. A study of radiation-induced dielectric breakdown has revealed the possibility of anomalous hot-electron injection to an excess of bulk oxide traps in the ionization channel produced by very heavily ionizing radiation, which leads to intrinsic breakdown in high-field stressed devices. This is interpreted in terms of a modified model for radiation-induced dielectric breakdown based upon the primary dependence of breakdown on charge injection rather than high-field mechanisms. A detailed investigation of charge trapping and interface state generation due to various radiations has revealed evidence of neutron induced interface states, and the generation of positive oxide charge in devices due to all the radiations tested. The greater the linear energy transfer of the radiation, the greater the magnitude of charge trapped in the oxide and the number of interface states generated. This is interpreted in terms of Si-H and Si-OH bond-breaking at the Si-SiO₂ interface which is enhanced by charge carrier transfer to the interface and by anomalous charge injection to compensate for the excess of charge carriers created by the radiation. (author)

[en] The main objective of the Depleted UO₂ Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO₂ kernels with diameters of 500 ± 20 (micro)m and 3.5 kg of UO₂ kernels with diameters of 350 ± 10 (micro)m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO₂ kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO₃ · 2H₂O microspheres to form dense UO₂ kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 ± 10-(micro)m-diameter kernels, and to obtain very high yields

[en] A spheroidal composite inorganic sorbent was developed for U.S. Department of Energy-Efficient Separations and Processing Crosscutting Program (USDOE-ESP) for potential use in removing radioactive cesium isotopes from acidic high-salt waste streams such as those at Idaho National Engineering and Environmental Laboratory (INEEL). The sorbent, zirconium monohydrogen phosphate (ZrHP) embedded with fine powder of ammonium molybdophosphate (AMP), was prepared using a unique internal gelation process which can be used to make porous reproducible microspheres that are structurally strong, have a low tendency for surface erosion, and improve the flow dynamics for column operations. Both ZrHP and AMP are excellent sorbent materials and, being inorganic, are stable in high radiation fields. AMP is a very effective sorbent for removing cesium from salt-bearing waste streams for a wide range of acidity. In the pH range of 2 to 10, ZrHP is also a very effective sorbent for removing Cs, Sr, Th, U(VI), Pu(IV), Am(III), Hg, and Pb from streams of lower ionic concentrations. Crucial to developing the spheroidal AMP-ZrHP sorbent was to determine the ideal weight percentage of AMP that could be embedded in the ZrHP microspheres in order to maintain the structural integrity of the microspheres and also achieve a good cesium separation. A total of 12 preparations were made. The dry weight percentage of AMP ranged from 30 to 60. Overall, the best composite microspheres prepared contained 50% AMP (by dry weight measurement). Another composite microsphere, which was composed of titanium monohydrogen phosphate (TiHP) embedded with 18 wt % (air-dried weight) potassium cobalt hexacyanoferrate (KCoCF) and developed for a different separations application, was also batch tested for comparison. It proved to be as effective in removing,the cesium as the air-dried AMP (50 wt %)-ZrHP. Granular KCoCF was also prepared and was very effective. Large samples of each of these materials were sent to INEEL for small-column testing with real waste

[en] The general objectives of this task are to develop, prepare, and test spheroidal inorganic ion exchangers made by the HMTA (hexamethylenetetramine) internal gelation process to remove radionuclides and heavy metals from waste streams occurring at the various DOE sites. Inorganic ion-exchange materials, such as sodium silicotitanate, sodium titanate, ammonium molybdeophosphate, phosphotungstic acid, hexacyanoferrates, titanium monohydrogen phosphate, hydrous titanium oxide, polyantimonic acid, magnesium oxide, etc. have high selectivities and efficiencies for separating and removing radionuclides (e.g., cesium, strontium, technetium, iodine, europium, cerium, ruthenium, and zirconium), actinides, and other elements (such as lead, mercury, silver, nickel, zinc, chromium, and fluoride) from aqueous waste streams. The development of cesium specific spherical sorbents for treatment of acidic, high-salt waste solutions was initiated in FY 1998. Acid-side treatment is important at INEEL and could become important if acidic sludge washing were to become a treatment option at Hanford, Savannah River, or Oak Ridge. Zirconium monohydrogen phosphates (ZrHP) embedded with ammonium molybdophosphate (AMP) was the cesium selective inorganic sorbent chosen for making microspheres. AMP is known to be a very effective sorbent for removing cesium from waste streams over a wide range of acidity and salinity, and it has very rapid loading kinetics. The cesium can also be eluted from AMP with ammonium salt solutions. AMP cannot be used as a sorbent at pHs above 7 because it decomposes. In the pH range of 1 to 7, ZrHP is also a very effective sorbent for removing Cs, Sr, Th, U(VI), Pu(IV), AM(III), Hg, and Pb from streams of lower ionic concentrations

[en] Approximately 50% of the fission product cesium was released from the overheated UO₂ fuel in the TMI-2 accident. Steam that boiled away from a water pool in the bottom of the reactor vessel transported the released fission products throughout the reactor coolant system (RCS). Some fission products passed directly through a leaking valve with steam and water into the containment structure, but most deposited on dry surfaces inside of the RCS before being dissolved or resuspended when the RCS was refilled with water. A cesium transport model was developed that extended measured cesium in the RCS back to the first day of the accident. The model revealed that ∼62% of the released ¹³⁷Cs deposited on dry surfaces inside of the RCS before being slowly leached and transported out of the RCS in leaked or letdown water. The leach rates from the model agreed reasonably well with those measured in the laboratory. The chemical behavior of cesium in the TMI-2 accident agreed with that observed in fission product release tests at Oak Ridge National Laboratory (ORNL)

[en] One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required

[en] Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules

[en] Experiments conducted at Oak Ridge National Laboratory both with fission product simulants and with irradiated commercial fuel have been utilized to develop a semi-empirical model of fission product release from defected Light Water Reactor (LWR) fuel rods. At fuel temperatures less than 1200⁰C, releases occur from fission products previously accumulated in the pellet-to-cladding gap region. In this temperature range, the release of species of moderate volatility is postulated to result from two processes. The first of these, which occurs during the period of fuel clad rupture, is due to the transport of the fill and fission product gases as they are vented through the cladding defect. The second mechanism for release, which is time-dependent, involves the diffusional transport of the semi-volatile species to the point of clad rupture through the interconnected voids (the pellet-to-cladding gap and cracks in fuel pellets) within the fuel rod