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[en] Between 1944 and 1989, the Hanford Site produced 60 percent (54.5 metric tons) of the United States weapons plutonium and produced an additional 12.9 metric tons of fuels-grade plutonium. High activity wastes, including plutonium lost from the separations processes used to isolate the plutonium, were discharged to underground storage tanks during these operations. Plutonium in the Hanford tank farms is estimated to be ∼700 kg but may be up to ∼1000 kg. Despite these apparent large quantities, the average plutonium concentration in the ∼200 million liter tank waste volume is only about 0.003 grams per liter (∼0.0002 wt%). The plutonium is largely associated with low solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO2 · xH2O, could undergo sufficient crystal growth through Ostwald ripening in the alkaline tank waste to potentially be separable from neutron absorbing constituents by settling or sedimentation. It was found that plutonium that entered the alkaline tank waste by precipitation through neutralization from acid solution is initially present as 2- to 3-nm (0.002- to 0.003-μm) scale PuO2 · xH2O crystallite particles and grows from that point at exceedingly slow rates, posing no risk to physical segregation. These conclusions are reached by both general considerations of Ostwald ripening and specific observations of the behaviors of PuO2 and PuO2 · xH2O upon aging in alkaline solution.
[en] Complex, high ionic strength media are used throughout the plutonium cycle, from its processing and purification in nitric acid, to waste storage and processing in alkaline solutions of concentrated electrolytes, to geologic disposal in brines. Plutonium oxidation/reduction, stability, radiolysis, solution and solid phase chemistry have been studied in such systems. In some cases, predictive models for describing Pu chemistry under such non-ideal conditions have been developed, which are usually based on empirical databases describing specific ion interactions. In Chapter 11, Non-Ideal Systems, studies on the behavior of Pu in various complex media and available model descriptions are reviewed
[en] K Basin sludge contains metallic uranium and uranium oxides that will corrode and hydrate during storage. The end-state (final) corrosion products will have a lower particle density and a higher void fraction (or volume fraction of sludge occupied by water) than the starting-state sludge at the beginning of storage. As the particle density and void fraction change, the volume occupied by a given mass of sludge will also change. The purpose of this report is to quantify how the various types and sources of K Basin sludge will react and volumetrically expand (or contract) between the time the sludge is first loaded into the storage containers (starting state) and the time all major volume-changing reactions have been completed (end state). The results from this report will be used in design and safety basis calculations for sludge management systems and will be incorporated into the sludge technical basis documents
[en] Processing of non-irradiated plutonium oxide, PuO2, at the Hanford Site has been done at the Plutonium Finishing Plant (PFP) and in recycle of PuO2-bearing fuels through Hanford reprocessing plants. Plutonium oxide is notoriously refractory and difficult to dissolve. As such, losses of PuO2 residues from the PFP or from Hanford reprocessing plants can report to Hanford Site underground tank waste storage. Those stored wastes are destined to become feed to the Waste Treatment Plant, WTP. Information on the processing of non-PFP plutonium oxide in Hanford plants is provided in this brief report. To help gain perspective, information on PFP processing and plutonium additions to the tank farm system from other Hanford processing is also presented. Processing of non-irradiated plutonium oxide, PuO2, scrap for recovery of plutonium values occurred routinely at Hanford's Plutonium Finishing Plant (PFP) in glovebox line operations. Plutonium oxide is difficult to dissolve, particularly if it has been high-fired; i.e., calcined to temperatures above about 400 C and much of it was. Dissolution of the PuO2 in the scrap typically was performed in PFP's Miscellaneous Treatment line using nitric acid (HNO3) containing some source of fluoride ion, F-, such as hydrofluoric acid (HF), sodium fluoride (NaF), or calcium fluoride (CaF2). The HNO3 concentration generally was 6 M or higher whereas the fluoride concentration was ∼0.5 M or lower. At higher fluoride concentrations, plutonium fluoride (PuF4) would precipitate, thus limiting the plutonium dissolution. Some plutonium-bearing scrap also contained PuF4 and thus required no added fluoride. Once the plutonium scrap was dissolved, the excess fluoride was complexed with aluminum ion, Al3+, added as aluminum nitrate, Al(NO3)3·9H2O, to limit collateral damage to the process equipment by the corrosive fluoride. Aluminum nitrate also was added in low quantities in processing PuF4. The PuO2 dissolution was not perfect, however, and some amount of PuO2 survived the HNO3/F- treatment, continued as solids through the solvent extraction processes at the PFP, and reported to the waste that was sent to the Hanford tank farms. The process raffinate disposed from PFP to tank farms largely went to tank 241-SY-102. Identification and characterization of plutonium-bearing solids, including PuO2, has been done (Callaway 2004). Not all PuO2 dissolution for plutonium recovery at Hanford occurred at PFP, however. If the PuO2 had been irradiated in a reactor, usually as part of a PuO2/UO2 (plutonium-uranium dioxide mixed oxide or MOX) fuel, the attendant high radiological doses precluded glovebox processing. Instead, the irradiated MOX with contained PuO2 would enter a Hanford reprocessing plant for dissolution and recovery of plutonium and also uranium values.
[en] This report summarizes information and data on the reaction behavior of uranium metal in water, in water-saturated simulated and genuine K Basin sludge, and in grout matrices. This information and data are used to establish the technical basis for metallic uranium reaction behavior for the K Basin Sludge Treatment Project (STP). The specific objective of this report is to consolidate the various sources of information into a concise document to serve as a high-level reference and road map for customers, regulators, and interested parties outside the STP (e.g., external reviewers, other DOE sites) to clearly understand the current basis for the corrosion of uranium metal in water, sludge, and grout
[en] The Westinghouse team has extended the Lawrence Livermore National Laboratory advanced conceptual design for the TPX PF magnets through preliminary design. This is the first time superconducting PF magnets have been designed for application in a tokamak. Particular challenges were encountered and solved in developing the coil insulation system, welding the helium stubs, and winding the coil. The authors fabricated a coil using copper stranded CIC conductor, to surface manufacturability issues and demonstrate the solutions
[en] Under contract to Fluor Hanford, Pacific Northwest National Laboratory directed laboratory, bench-scale, and pilot-scale vendor testing to evaluate the use of commercial gravity mineral concentration technology to remove and concentrate uranium metal from Hanford K Basin sludge. Uranium metal in the sludge corrodes by reacting with water to generate heat and hydrogen gas, and may constrain shipment and disposal of the sludge to the Waste Isolation Pilot Plant as remote-handled transuranic waste. Separating uranium metal from the K Basin sludge is expected to be similar to some gold recovery operations. Consequently, the capabilities of commercial gravity mineral concentration technologies were assessed for their applicability to K Basin sludge streams. Overall, the vendor testing demonstrated the technical feasibility of using gravity concentration equipment to separate the K Basin sludge into a high-volume uranium metal-depleted stream and a low-volume uranium metal-rich stream. I n test systems, more than 96% of the uranium metal surrogate was concentrated into 10 to 30% of the sludge mass (7 to 24% of the sludge volume). With more prototypical equipment and stream recycle, higher recoveries may be achieved
[en] This test plan was prepared by the Pacific Northwest National Laboratory (PNNL) under contract with Fluor Hanford (FH). The test plan describes the scope and conditions to be used to perform laboratory-scale testing of the Sludge Treatment Project (STP) hydrothermal treatment of K Basin sludge. The STP, managed for the U. S. Department of Energy (DOE) by FH, was created to design and operate a process to eliminate uranium metal from the sludge prior to packaging for Waste Isolation Pilot Plant (WIPP) by using high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. The proposed testing builds on the approach and laboratory test findings for both K Basin sludge and simulated sludge garnered during prior testing from September 2006 to March 2007. The outlined testing in this plan is designed to yield further understanding of the nature of the chemical reactions, the effects of compositional and process variations and the effectiveness of various strategies to mitigate the observed high shear strength phenomenon observed during the prior testing. These tests are designed to provide process validation and refinement vs. process development and design input. The expected outcome is to establish a level of understanding of the chemistry such that successful operating strategies and parameters can be implemented within the confines of the existing STP corrosion vessel design. In July 2007, the DOE provided direction to FH regarding significant changes to the scope of the overall STP. As a result of the changes, FH directed PNNL to stop work on most of the planned activities covered in this test plan. Therefore, it is unlikely the testing described here will be performed. However, to preserve the test strategy and details developed to date, the test plan has been published
[en] Means to decrease the rate of hydrogen gas generation from the chemical reaction of uranium metal with water were identified by surveying the technical literature. The underlying chemistry and potential side reactions were explored by conducting 61 principal experiments. Several methods achieved significant hydrogen gas generation rate mitigation. Gas-generating side reactions from interactions of organics or sludge constituents with mitigating agents were observed. Further testing is recommended to develop deeper knowledge of the underlying chemistry and to advance the technology aturation level. Uranium metal reacts with water in K Basin sludge to form uranium hydride (UH3), uranium dioxide or uraninite (UO2), and diatomic hydrogen (H2). Mechanistic studies show that hydrogen radicals (H·) and UH3 serve as intermediates in the reaction of uranium metal with water to produce H2 and UO2. Because H2 is flammable, its release into the gas phase above K Basin sludge during sludge storage, processing, immobilization, shipment, and disposal is a concern to the safety of those operations. Findings from the technical literature and from experimental investigations with simple chemical systems (including uranium metal in water), in the presence of individual sludge simulant components, with complete sludge simulants, and with actual K Basin sludge are presented in this report. Based on the literature review and intermediate lab test results, sodium nitrate, sodium nitrite, Nochar Acid Bond N960, disodium hydrogen phosphate, and hexavalent uranium [U(VI)] were tested for their effects in decreasing the rate of hydrogen generation from the reaction of uranium metal with water. Nitrate and nitrite each were effective, decreasing hydrogen generation rates in actual sludge by factors of about 100 to 1000 when used at 0.5 molar (M) concentrations. Higher attenuation factors were achieved in tests with aqueous solutions alone. Nochar N960, a water sorbent, decreased hydrogen generation by no more than a factor of three while disodium phosphate increased the corrosion and hydrogen generation rates slightly. U(VI) showed some promise in attenuating hydrogen but only initial testing was completed. Uranium metal corrosion rates also were measured. Under many conditions showing high hydrogen gas attenuation, uranium metal continued to corrode at rates approaching those observed without additives. This combination of high hydrogen attenuation with relatively unabated uranium metal corrosion is significant as it provides a means to eliminate uranium metal by its corrosion in water without the accompanying hazards otherwise presented by hydrogen generation.