Process Chemistry and Engineering Posters

Solvent Extraction Process Development for Partitioning and Transmutation of Spent Fuel

Monica C. Regalbuto, Jacqueline M. Copple, Ralph Leonard, Candid.Pereira, and Georg.F.Vandegrif.

Argonne National Laboratory, along with other national laboratories, has been developing a solvent extraction process for partitioning of spent fuel constituents to lead to safer and cheaper disposal of high-level waste. The process, known as UREX+ separates key radionuclides from dissolved spent fuel into

1. Uranium for disposal as LLW
2. Technetium for disposal as HLW
3. Iodine for disposal as HLW
4. Cs/Sr for decay storage followed by disposal as LLW
5. Pu/Np for production of mixed-oxide fuel
6. Am/Cm for fast-reactor fuel or transmutation
7. All remaining soluble fission products for disposal.

This poster discusses the Argonne development effort including

1. Design of process flowsheets by data collection and flowsheet development using the AMUSE code
2. Design of an advanced centrifugal contactor
3. Demonstration of alternative solvent extraction flowsheet options for the UREX+ process using dissolved fuel.

Research funded by U.S. Department of Energy, Office of Nuclear Energy, Science and Technology.

Contact Candido Pereira (2-9832, pereira@cmt.anl.gov). View Poster

Dissolution and Analysis of Nuclear Fuels and Targets

Allen Bakel and Artem Gelis

Two separate but related projects are briefly described

1. Dissolution of an irradiated commercial oxide nuclear fuel under high-temperature and low-acid conditions
2. Dissolution of uranium metal targets using permanganate as an oxidant.

A pin of irradiated Big Rock Point uranium oxide fuel was dissolved in nitric acid at elevated temperature to provide feedstock for the UREX+ demonstration. Elevated temperature was used to promote the complete dissolution of noble metals at relatively low nitric acid concentrations. Three products were obtained from this dissolution:

1. A dissolved fuel solution
2. Undissolved residue that was filtered from the solution
3. Leached cladding that contained no observable undissolved fuel.

Complete elemental analyses of the dissolved fuel, and residue are presented here. The data presented show that 99% of the fuel, including the noble metals was dissolved directly into the acid. The small amount of residue contained primarily Zr, Mo, and Pu. A Pu-bearing zirconium molybdate phase is commonly seen as a precipitate from dissolved fuels. While the total amount of residue is small, approximately 20% of the total Pu was found in the residue. Foils of depleted irradiated uranium metal, proposed as targets in ⁹⁹Mo production, were dissolved at elevated temperature and pressure. A stirred digester was used for the digestion of the low burn-up targets in KMnO₄ alkaline media. This procedure is novel in that it uses KMnO₄ to oxidize the U metal to uranyl hydroxide, increasing the Mo release into the solution.

This research is funded by the U.S. Department of Energy, Office of Nuclear Nonproliferation.

Contact Allen Bakel (630-252-9825, bakel@cmt.anl.gov).View Poster

Evaluation of Tris-CMPO and Tris-CMPS Ligands as Extractants for Actinide/Lanthanide Separation

Artem Guelis, Monica C. Regalbuto, George Vandegrift, Kornelia Matloka,^a Ajay K. Sah,^a Mathew W. Peters,^a Priya Srinivasan,^a and Michael J. Scott^a
aDepartment of Chemistry, University of Florida

Efficient protocols for the separation of actinide and lanthanide ions are required for the processing of acidic nuclear waste streams. Because actinide and lanthanide ions are of similar size and charge, the development of methods to selectively bind actinide(III) ions in the presence of lanthanide(III) metals has proven to be especially problematic. The successful separation method for actinides over trivalent lanthanides must exploit the slight differences in ionic radii and covalency/polarizability of these metal ions. A new group of the organic ligands was developed at the University of Florida to meet the criteria. A collaboration between the Chemical Engineering Division and the University of Florida is focusing on developing the solvent extraction process for Am/Ln separation in nuclear waste.

Research funded by U.S. Department of Energy, Office of Nuclear Energy, Science and Technology, Nuclear Energy Research Initiative (NERI).

Contact Art Guelis (630-252-6057, guelis@cmt.anl.gov). View Poster

Deliquescence of Fission Product Salts: Implications for the Corrosion Behavior of Spent Fuel

James Jerden

To design safe storage and disposal systems for spent nuclear fuel it is essential to have a mechanistic understanding of its corrosion behavior. It is particularly important to understand under what conditions (thermal, humidity) liquid water will contact damaged fuel pins. This is essential because the corrosion of spent fuel in aqueous solutions could lead to the leaching and release of radionuclides into the surrounding environment. When spent fuel is exposed to humid air an aqueous solution may accumulate on its surface due to the presence of deliquescent minerals. These minerals, which form in the reactor due to the concentration of volatile fission products such as cesium, draw water from humid air to form a saturated solution (brine) that may initiate the spent fuel corrosion process. We are performing experiments focused on elucidating how the deliquescence of fission product minerals may initiate and control the corrosion behavior of spent nuclear fuel.

Research funded by DOE Office of Civilian Radioactive Waste Management, Science and Technology Program.

Contact Jim Jerden (630-252-6497, jerden@cmt.anl.gov). View Poster

Surface Complexation of Neptunium with Iron Oxides

James Jerden and Jeremy Kropf

Neptunium is the element of primary concern for the long-term (>70,000 years) performance of the proposed nuclear waste repository at Yucca Mountain, Nevada. Predictive modeling of the release of neptunium from the waste canisters and its geosphere-biosphere transport requires a mechanistic understanding of how it interacts with solid surfaces (e.g., corroded steel, geologic media) along its release pathway. Our experiments focus on describing the molecular structures of the surface complexes that neptunium forms when it interacts with relevant mineral surfaces (e.g., iron oxides) under different thermal and chemical conditions. The main analytical technique we are using is x-ray absorption fine structure performed at the Advanced Photon Source.

Research funded by DOE Office of Civilian Radioactive Waste Management, Science and Technology Program.

Contact Jim Jerden (630-252-6497, jerden@cmt.anl.gov). View poster

Crystal Chemistry of Radionuclides in Spent Nuclear Fuel and Its Alteration Products

Jeffrey A. Fortner, A. Jeremy Kropf, Robert J. Finch and James C. Cunnane

Spent uranium oxide nuclear fuel hosts a variety of trace impurities, forged by nuclear fission, neutron capture, and radioactive decay. In the high-temperature, high-radiation field environment of a reactor (and, to a lesser degree, during subsequent storage) these trace elements are either incorporated into the fluorite structure of the uranium dioxide (e.g., neptunium, plutonium) or exsolved into separate phases (e.g., technetium, molybdenum, and ruthenium). We use synchrotron x-ray absorption spectroscopy and microscopy to reveal the local environment and chemistry of neptunium, plutonium, technetium, and molybdenum within spent uranium oxide nuclear fuel. Obtaining useful x-ray spectroscopic information requires that the background from the matrix uranium be substantially decreased, which we achieved using a bent-Laue diffractive optics energy analyzer with a bandwidth of about 75 eV. We find the plutonium and neptunium in spent fuel are tetravalent, with each having a local environment consistent with solid-solution in the uranium dioxide matrix. The majority of molybdenum and technetium were found to exist together and to have identical local environments, consistent with metallic epsilon-ruthenium. These findings hold even for some fuel particles that have been partially oxidized by aqueous corrosion at 90 C in air. Current data for the standard reduction potentials of the redox couples NpO₂⁺/Np⁴⁺ and UO₂²⁺/U⁴⁺ suggest that the UO₂²⁺ /U⁴⁺ couple may keep the corrosion potential below that required to oxidize Np(IV) to Np(V) as long as U(IV) predominates (that is, until most of the U(IV) has been oxidized).

This research is funded by the U.S. Department of Energy, Office of Civilian Radioactive Waste, Office of Science and Technology and International.

Contact Jeff Fortner (630-252-5594, fortner@cmt.anl.gov). View Poster .