Chuck Z. Soderquist

Corrosion of commercial spent nuclear fuel. 2. Radiochemical analyses of metastudtite and leachates

Summary Immersing commercial spent nuclear fuel (CSNF) in deionized water produced two corrosion products after a 2-year contact period. Suspensions of aggregates were observed to form at the air–water interface for each of five samples. These suspended aggregates were characterized by X-ray diffraction (XRD) to be metastudtite (UO4·2H2O), while the corrosion present on the surface of the fuel itself was determined to be studtite [(UO2)(O2)(H2O)2](H2O)2]. The presence of unreacted UO2 matrix was below the limits of detection by XRD for the three samples examined. The result prompted a radiochemical analysis of the solids collected from the sample air–water interface. The analysis indicated that high concentrations of 90Sr, 137Cs, and 99Tc, relative to the fuel inventory, had concentrated at the air–water interface along with the aggregates of metastudtite. Concentrations of 241Am were at least two orders of magnitude lower than expected in these solids, and retention of 237Np and 239Pu into the corrosion product was observed. The combined radiochemical analyses of the air–water interface aggregates and leachate samples are a rare example of radionuclide partitioning to an alteration phase and may provide preliminary evidence for mechanisms that give rise to such noticeable departures from fuel-inventory values. The leachate radiochemical data are compared to existing data from hydration of the same CSNF.

Tc Reductant Chemistry and Crucible Melting Studies with Simulated Hanford Low-Activity Waste

The FY 2003 risk assessment (RA) of bulk vitrification (BV) waste packages used 0.3 wt% of the technetium (Tc) inventory as a leachable salt and found it sufficient to create a significant peak in the groundwater concentration in a 100-meter down-gradient well. Although this peak met regulatory limits, considering uncertainty in the actual Tc salt fraction, peak concentrations could exceed the maximum concentration limit (MCL) under some scenarios so reducing the leachable salt inventory is desirable. The main objective of this study was to reduce the mobile Tc species available within a BV disposal package by reducing the oxidation state of the Tc in the waste feed and/or during melting because Tc in its reduced form of Tc(IV) has a much lower volatility than Tc(VII). Reduced Tc volatility has a secondary benefit of increasing the Tc retention in glass.

Probing the oxygen environment in UO22+ by solid-state O17 nuclear magnetic resonance spectroscopy and relativistic density functional calculations

A combined theoretical and solid-state (17)O nuclear magnetic resonance (NMR) study of the electronic structure of the uranyl ion UO(2)(2+) in (NH(4))(4)UO(2)(CO(3))(3) and rutherfordine (UO(2)CO(3)) is presented, the former representing a system with a hydrogen-bonding environment around the uranyl oxygens and the latter exemplifying a uranyl environment without hydrogens. Relativistic density functional calculations reveal unique features of the U-O covalent bond, including the finding of (17)O chemical shift anisotropies that are among the largest for oxygen ever reported (>1200 ppm). Computational results for the oxygen electric field gradient tensor are found to be consistently larger in magnitude than experimental solid-state (17)O NMR measurements in a 7.05 T magnetic field indicate. A modified version of the Solomon theory of the two-spin echo amplitude for a spin-5/2 nucleus is developed and applied to the analysis of the (17)O echo signal of U (17)O(2)(2+).

Caustic Leaching of Hanford Tank S-110 Sludge

This report describes the Hanford Tank S-110 sludge caustic leaching test conducted in FY 2001 at the Pacific Northwest National Laboratory. The data presented here can be used to develop the baseline and alternative flowsheets for pretreating Hanford tank sludge. The U.S. Department of Energy funded the work through the Efficient Separations and Processing Crosscutting Program (ESP; EM﷓50).

Production of High-purity Radium-223 from Legacy Actinium-Beryllium Neutron Sources

Radium-223 is a short-lived alpha-particle-emitting radionuclide with potential applications in cancer treatment. Research to develop new radiopharmaceuticals employing (223)Ra has been hindered by poor availability due to the small quantities of parent actinium-227 available world-wide. The purpose of this study was to develop innovative and cost-effective methods to obtain high-purity (223)Ra from (227)Ac. We obtained (227)Ac from two surplus actinium-beryllium neutron generators. We retrieved the actinium/beryllium buttons from the sources and dissolved them in a sulfuric-nitric acid solution. A crude actinium solid was recovered from the solution by coprecipitation with thorium fluoride, leaving beryllium in solution. The crude actinium was purified to provide about 40 milligrams of actinium nitrate using anion exchange in methanol-water-nitric acid solution. The purified actinium was then used to generate high-purity (223)Ra. We extracted (223)Ra using anion exchange in a methanol-water-nitric acid solution. After the radium was separated, actinium and thorium were then eluted from the column and dried for interim storage. This single-pass separation produces high purity, carrier-free (223)Ra product, and does not disturb the (227)Ac/(227)Th equilibrium. A high purity, carrier-free (227)Th was also obtained from the actinium using a similar anion exchange in nitric acid. These methods enable efficient production of (223)Ra for research and new alpha-emitter radiopharmaceutical development.

Chemical Trends in Solid Alkali Pertechnetates

Insight into the solid-state chemistry of pure technetium-99 (99Tc) oxides is required in the development of a robust immobilization and disposal system for nuclear waste stemming from the radiopharmaceutical industry, from the production of nuclear weapons, and from spent nuclear fuel. However, because of its radiotoxicity and the subsequent requirement of special facilities and handling procedures for research, only a few studies have been completed, many of which are over 20 years old. In this study, we report the synthesis of pure alkali pertechnetates (sodium, potassium, rubidium, and cesium) and analysis of these compounds by Raman spectroscopy, X-ray absorption spectroscopy (XANES and EXAFS), solid-state nuclear magnetic resonance (static and magic angle spinning), and neutron diffraction. The structures and spectral signatures of these compounds will aid in refining the understanding of 99Tc incorporation into and release from nuclear waste glasses. NaTcO4 shows aspects of the relatively higher electronegativity of the Na atom, resulting in large distortions of the pertechnetate tetrahedron and deshielding of the 99Tc nucleus relative to the aqueous TcO4-. At the other extreme, the large Cs and Rb atoms interact only weakly with the pertechnetate, have closer to perfect tetrahedral symmetry at the Tc atom, and have very similar vibrational spectra, even though the crystal structure of CsTcO4 is orthorhombic while that of RbTcO4 is tetragonal. Further trends are observed in the cell volume and quadrupolar coupling constant.

A Radiochemical Analyses of Metastudtite and Leachates from Spent Fuel

Immersion of commercial spent nuclear fuel (CSNF) in deionized water produced two novel corrosion products after a two-year contact period. Another unexpected result was that suspensions of aggregates were observed to form at the air-water interface for each of five samples. These solids were characterized, by SEM and XRD to be nearly pure metastudtite (UO4-2H2O); while the corrosion present on the surface of the fuel itself was determined to be studtite (UO4-2H2O). The occurrence of the floating phase prompted a radiochemical analysis of these solids. This chemical analysis was a unique opportunity to study the relatively pure corrosion phase for incorporation of radionuclides. The analysis indicated that high concentration of 90Sr, 137Cs, 99Tc, and that lower concentrations 237Np, 238, 239Pu and 243, 244Cm had partitioned with the air-water interface aggregates. The concentrations of 241Am were two orders of magnitude lower than the expected inventory in the suspended solids. The radiochemical analyses of the several leachate samples provide preliminary solubility data for the hydrogen peroxide leaching of CSNF and these data are compared to leaching of the same fuel in J-13 and deionized waters. The extent of fuel dissolution in these media are discussed.

Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis

Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO2 fuel dissolved in nitric acid and UO2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared.

A non-aqueous reduction process for purifying 153Gd produced in natural europium targets

Gadolinium-153 is a low-energy gamma-emitter used in nuclear medicine imaging quality assurance. Produced in nuclear reactors using natural Eu₂O₃ targets, ¹⁵³Gd is radiochemically separated from europium isotopes by europium reduction. However, conventional aqueous europium reduction produces hydrogen gas, a flammability hazard in radiological hot cells. We altered the traditional reduction method, using methanol as the process solvent to nearly eliminate hydrogen gas production. This new, non-aqueous reduction process demonstrates greater than 98% europium removal and gadolinium yields of 90%.

Initial Results From Dissolution Testing of Spent Fuel Under Acidic Conditions

Flow through dissolution tests using leachants with pH in the range 2 to 7 have been conducted on a moderate burnup Light Water Reactor spent fuel. Such low pH conditions have been modeled as possibly occurring in a failed waste package at the proposed repository at Yucca Mountain. The release of total uranium, 99Tc, 90Sr, 137Cs, and 239&240Pu were measured for up to 90% total reaction of the specimens. The reaction rates, determine both from the cumulative release and the release normalized to surface area, were found to decrease with increasing pH and with increasing extent of reaction. The implications to instantaneous release and long-term behavior in a geologic repository are discussed.