Nuclear Waste Streams Research Articles

Evaluation of an intermediate-silica sedimentary chabazite as exchanger for potentially radioactive cations

Intermediate-silica sedimentary chabazite contained in two chabazite-rich tuffaceous rocks has been evaluated as potential cation exchanger for radioactive cation removal from nuclear waste streams. Exchange isotherms have been obtained for the cationic couples Na/Ba, Na/Co, Na/Cs and Na/Sr and the relevant thermodynamic parameters calculated with the help of a computer program. Sedimentary chabazite turns out very selective for Cs +, fairly selective for Ba 2+ and Sr 2+, especially at low equivalent fraction of these cations in solution, and unselective for Co 2+. An explanation of the observed selectivities is attempted on the basis of the structural features of chabazite.

An Investigation into the Oxidation State of Molybdenum in Simplifies High Level Nuclear Waste Glass Compositions

The redox state of Mo in glasses containing simplified simulated high level nuclear waste (HLW) streams has been investigated using Electron Spin Resonance (ESR) and X-ray Photoelectron Spectroscopy (XPS). Melts performed in highly oxidising or neutral (air) atmospheres contained Mo6+, but our study indicates that the Mo can be at least partially reduced to Mo4+ or Mo3+ by melting in a reducing atmosphere. The implications for glasses containing a full simulated HLW stream are that the formation of detrimental crystalline phases containing Mo6+ upon heat treatment may be avoided by reducing the oxidation state of Mo during melting.

U.S. accelerator-driven transmutation of waste (ATW) program objectives, and technology development efforts

The U.S. Program to evaluate accelerator-driven systems for transmuting problematic, long-lived nuclear waste stream components was initiated during fiscal year 2000, based largely on the Accelerator-driven Transmutation of Waste (ATW) Technology Development Road Map developed during 1999. The Road Map (DOE/RW-0519) effort provided a long-range plan, involving technology development, demonstration, and deployment, as well as a recommended initial effort to evaluate the technology options for five or six years. This paper summarizes the ATW Research and Development Plan currently in draft form. Technology Readiness Levels (TRLs), which are based in part on a system used by the U.S. National Aeronautics and Space Administration in determining levels of flight readiness, was developed for use in assessing and advancing technologies relevant to waste transmutation. Based on TRLs and other considerations, the Program is screening technology options and prioritizing the long-term research and development effort. A top-level schedule illustrates the efforts planned to advance the important technology options in preparation for integrated system tests.

Enhanced selectivity for actinides over lanthanides with CMPO ligands secured to a C(3)-symmetric triphenoxymethane platform.

A ligand system containing three preorganized carbamoylmethylphosphine oxide (CMPO) moieties anchored onto a rigid C(3)-symmetric triphenoxymethane platform has been developed for facile metal complexation and subsequent extraction from aqueous acidic nuclear waste streams. Intended to mimic the 3:1 CMPO-actinide stoichiometry of the extracted species in the TRUEX nuclear waste treatment process, the CMPO arms on this ligand are oriented such that all three CMPO moieties can cooperatively bind a metal ion. Extractions of simulated nuclear waste streams (10(-4) M metal in 1 M nitric acid) with solutions of this ligand in methylene chloride (10(-3) M) reveal a high affinity for the actinide thorium and a very low, but constant, affinity for the lanthanides across the series. Thorium and five lanthanide (lanthanum, cerium, neodymium, europium, and ytterbium) nitrate complexes of this ligand have been synthesized and fully characterized by X-ray crystallography, (1)H and (31)P NMR spectra, and FT-ICR-MS to elucidate the mechanism of this unique actinide selectivity. All six oxygen donors from the three CMPO arms of the ligand and one or two nitrate counterions coordinate these metals to afford 2+ cationic complexes in every case. Because of the large size of the ligand, both the thorium and lanthanide complexes present similarly charged and sized surfaces to the extraction solvents, but the thorium complex is extracted quantitatively over the lanthanide complexes. A possible rationale for this extraction behavior difference is presented and further illustrated by the extraction properties of this ligand system for the alkali metals (lithium, sodium, potassium, rubidium, and cesium) as picrate salts and by the solid- and solution-state structures of its lithium picrate complex.

Extraction of heptavalent technetium from nuclear waste streams with 2-nitrophenyl octyl ether

Technetium-99, in the form of Tc(VII), is one of major problems in the management and disposal of nuclear waste in view of its long half-life in combination with high geochemical mobility. Solvent extraction is an important separation technique to remove99Tc from liquid nuclear waste streams. 2-nitrophenyl octyl ether (NPOE) shows high extraction ability for Tc(VII) from water and NaOH solutions (>99.5%) in spite of the lack of donating atoms. Upon extraction from alkaline solutions, TcO4-associates with NPOE molecules in 1:1 stoichiometry where the co-extracted cation is expected to play an important role in the extraction mechanism. At pH > 4 the Tc(VII) distribution is relatively insensitive towards changes in pH, while at decreasing pH (<4) the distribution ratio decreases sharply. The strong dependency of the Tc(VII) distribution on the acidity is very useful for developing efficient extraction-stripping systems. This study showed that NPOE is an excellent compound for the removal of technetium from alkaline nuclear wastes.

Extraction of lanthanides and actinides by a magnetically assisted chemical separation technique based on CMPO-calix[4]arenes

A novel particulate system carrying CMPO ligands pre-organised on a calixarene scaffold has been synthesised and demonstrated to extract Eu3+, Am3+ and Ce3+ at high efficiency from simulated nuclear waste streams.

Membrane technologies for liquid radioactive waste treatment

The paper deals with some problems concerning reduction of radioactivity of liquid low-level nuclear waste streams (LLLW). The membrane processes as ultrafiltration (UF), seeded ultrafiltration (SUF), reverse osmosis (RO) and membrane distillation (MD) were examined. Ultrafiltration enables the removal of particles with molecular weight above cut-off of UF membranes and can be only used as a pre-treatment stage. The improvement of removal is achieved by SUF, employing macromolecular ligands binding radioactive ions. The reduction of radioactivity in LLLW to very low level were achieved with RO membranes. The results of experiments led the authors to the design and construction of UF+2RO pilot plant. The development of membrane distillation improve the selectivity of membrane process in some cases. The possibility of utilisation of waste heat from cooling system of nuclear reactors as a preferable energy source can significantly reduce the cost of operation.

Fast source preparation for alpha-spectrometry of uranium and transuranium isotopes

A very fast and simple source preparation procedure for α-spectrometry, especially for the Large-Area Grid Ionization Chamber, has been developed. It is based on two liquid-liquid extractions and direct evaporation of the extractants, containing the uranium and transuranium isotopes, on the counting dishes. From the sample two thin-layer counting sources are prepared to take into account the α-spectrometric interference of the nuclides234U−237Np and238Pu−241Am, where one counting source contains the U- and Pu-isotopes and the other one the Np- and Am-isotopes. The chemical yields are in the range of 95 and 98% so that no yield tracers are necessary. Starting with the dissolved sample, the two counting sources can be prepared in only about 1 hour. A second even faster preparation version, with only one counting source, has been devised, if discrimination of the interfering nuclides mentioned above is dispensed with. This version can also be used for gross α-activity measurements in a liquid scintillation counter. With the preparation procedures described here the sample volume is limited to 1 ml. Therefore it is preferably applied to medium-active samples from nuclear power plant waste stream, but it can also be used for low-level environmental samples after suitable preconcentration procedures.

Radiochemical Stability of Dicyclohexano-18-Crown-6 Ether (DCH18C6) and its Use in a Recovery Process of Strontium from Acidic Nuclear Waste Stream

Article Radiochemical Stability of Dicyclohexano-18-Crown-6 Ether (DCH18C6) and its Use in a Recovery Process of Strontium from Acidic Nuclear Waste Stream was published on December 1, 1997 in the journal Radiochimica Acta (volume 78, issue s1).

A Recovery Process of Strontium from Acidic Nuclear Waste Streams

A process combining liquid–liquid extraction with ion exchange for strontium traces removal in the presence of an excess of sodium concentrations is described. In this continuous process, strontium is extracted from acidic (0.9 M) solution using a substoichiometric amount of dicyclohexano-18-crown-6 (0.1 M in CHCl3). Extracted strontium is readily stripped from an organic phase with water and concentrated onto an ion exchanger. This process allows a high strontium decontamination factor (250) with a very important extraction factor (5000).

Electrochemical reduction of nitrates and nitrites in alkaline nuclear waste solutions

Sodium nitrate and nitrite are major components of alkaline nuclear waste streams and contribute to environmental release hazards. The electrochemical reduction of these materials to gaseous products has been studied in a synthetic waste mixture. The effects of electrode materials, cell design, and other experimental parameters have been investigated. Lead was found to be the best cathode material in terms of current efficiency for the reduction of nitrate and nitrite in the synthetic mix. The current efficiency for nitrite and nitrate removal is improved in divided cells due to the elimination of anodic oxidation of nitrite. Operation of the divided cells at high current densities (300–600 mA cm−2) and high temperatures (80°C) provides more efficient reduction of nitrite and nitrate. Nearly complete reduction of nitrite and nitrate to nitrogen, ammonia, or nitrous oxide was demonstrated in 1000 h tests in a divided laboratory electrochemical flow cell using a lead cathode, Nafion® 417 cation exchange membrane, and oxygen evolving DSA® or platinum clad niobium anode at a current density of 500 mA cm−2 and a temperature of 70° C. Greater than 99% of the nitrite and nitrate was removed from the synthetic waste mix batch in the 1000 h tests at an overall destruction efficiency of 55%. The process developed shows promise for treating large volumes of waste.

Nuclear utility mixed waste stream characterization

Nuclear utility mixed waste stream characterization

Review of biotechnology applications to nuclear waste treatment.

This paper gives an overview of the feasibility of the application of biotechnology to nuclear waste treatment. The contents are based on a report which PA Technology carried out for the Department of the Environment (DoE Reference: DoE/RW/88.008 Sector No 2.3). Many living and dead organisms accumulate heavy metals and radionuclides. The controlled use of this phenomenon forms the basis for the application of biotechnology to the removal of radionuclides from nuclear waste streams. Indeed, biotechnology offers a series of new opportunities for removal of radionuclides from dilute aqueous process effluents. Such technology is already used for heavy metal removal on a commercial basis and could be optimised for radionuclide removal. An overview of biotechnology areas, namely the use of biopolymers and biosorption using biomass applicable to the removal of radionuclides from industrial nuclear effluents is given. The potential of biomagnetic separation technology, genetic engineering and monoclonal antibody technology is also to be examined. The most appropriate technologies to develop for radionuclide removal in the short term appear to be those based on biosorption of radionuclides by biomass and the use of modified and unmodified biopolymers in the medium term.

Nuclear waste ponds and streams on the Hanford Site: an ecological search for radiation effects.

Nuclear waste ponds and streams on the Hanford Site: an ecological search for radiation effects.

A Conceptual Reprocessing and Waste Management Scheme for Zircaloy-Clad Thoria Fuels

A family of inorganic ion exchangers has been under development in this laboratory for stabilization of nuclides from various nuclear processing waste streams.1 These materials are formed from hydrous oxides of titanium, zirconium, niobium, or tantalum and have a high affinity for all polyvalent cations in the wastes and a large ion exchange capacity. They consolidate and convert to nonvolatile, leach-resistant ceramic waste forms. While their effectiveness for high-level waste stabilization was demonstrated mainly with the titanium form of the ion exchanger, the zirconium and other forms can also be used. The large source of zirconium in the form of waste Zircaloy fuel hulls offers the particu- larly interesting possibility of combining this waste stream with others to mutual benefit. In fact, there is more than enough zirconium in this high-level solid waste to stabilize all the associated (e.g., Purex) wastes in the form of stable zirconates and to simultaneously reduce the resulting high- level volume by a factor of ∼2 over the traditional calcine- glass approach. This observation has been investigated, and a conceptual process to convert this Zircaloy waste to ion exchange material was recently described.2,3 This process basically involved distillation of ZrCl4 from the chlorinated waste, followed by reaction to form alkoxides, and then hydrolysis.