Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

et al. 2011

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ph: (865) 576-8401 fax: (865) 576 5728 email: reports@adonis.osti.gov Available to the public from the National Technical Information Service, U.S. Department of Commerce, 5285 Port Royal Rd., Springfield, VA 22161 ph: (800) 553-6847 fax: (703) 605-6900 email: orders@nits.fedworld.gov online ordering: http://www.ntis.gov/ordering.htm Abstract viiThese findings indicate that for solids very rich in uranium (≳70 to 80 wt%, dry basis), long-term storage under aerated non-agitated conditions will form a thin metaschoepite crust that is readily broken. The underlying material may compact somewhat with time, but, because little appreciable change in settled solids strength was seen after nearly a year of settling at 51°C in laboratory tests, little increase in strength, beyond that associated with any compaction, is anticipated during lower temperature but longer term storage.The observations from the 28-month settling test with uranium-rich genuine sludge and the studies with the uraninite oxidation suggest that metaschoepite crystal ripening and intergrowth may have been responsible for the uncommon strength found in the genuine sludge. In this case, the presence of relatively high amounts of metaschoepite within the settled solids layer likely was necessary to produce the strong product material. Based on these findings, K Basin sludges rich in metaschoepite [≳70 to 80 wt% U as U(VI), dry basis] may self-cement by Ostwald ripening to produce strong agglomerates similar to the behavior shown by the 96-13 sludge in the 28-month settling tests. Because of the low solubility of UO 2 , similar gains in strength by Ostwald ripening for UO 2 -rich sludge cannot occur.  Under 51°C oxygenated conditions, a full KW containerized sludge simulant containing gibbsite, ferrihydrite, mordenite, organic ion exchange resin, Hanford sand, and ~50%/50% UO 2 /UO 3 ·2H 2 O decreased in settled solids volume by about 25% over 106 days while a parallel control test run in the absence of oxygen decreased only about 3%. The volume decrease coincided with the oxidation of UO 2 to UO 3 ·2H 2 O and may have been due to better solids packing or coagulation of non-uranium solids with the crystallizing UO 3 ·2H 2 O. However, even though the KW containerized sludge simulant compacted with time, the strength remained low. Under warm (30±5ºC) semi-oxic conditions, the uraninite in uranium-rich K Basin sludge samples oxidized to metaschoepite and other U(VI) phases after 9 years of hot cell storage (Delegard et al. 2007a).These findings mean that, unlike uraninite-rich sludges that form continuous layers or networks of product metaschoepite upon reaction with oxygen or which already have high metaschoepite concentrations, sludges more dilute in uraninite or metaschoepite likely will be unable to produce continuous metaschoepite layers that inhibit further oxidation or which self-cement to form highstrength agglomerates. Instead, the metaschoepite may act to coagulate non-uranium sludge solids and produce settled solids that are more tightly packed ...

Section: Uranium Oxidesmentioning

confidence: 94%

Effects of Time, Heat, and Oxygen on K Basin Sludge Agglomeration, Strength, and Solids Volume

Delegard¹,

Sinkov²,

et al. 2011

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“…(a) The Na 2 SO 4 /H 3 PO 4 solution may be analyzed spectrophotometrically to determine the oxidation state distribution of the oxidized uranium compounds (Sinkov et al 2008) or may be discarded.…”

Section: Methods Summarymentioning

confidence: 99%

“…(Sinkov et al 2008). These summed dry weights were complemented by a matching weight of added water (i.e., the simulated sludge was 50 wt% water).…”

Section: Experiments For Uranium Metal Dissolution Rate and Analysis mentioning

confidence: 99%

Uranium Metal Analysis via Selective Dissolution

Delegard¹,

Sinkov²,

et al. 2008

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SummaryUranium metal, which is present in sludge held in the Hanford Site K West Basin, can create hazardous hydrogen atmospheres during sludge handling, immobilization, or subsequent transport and storage operations by its oxidation/corrosion in water. A thorough knowledge of the uranium metal concentration in sludge therefore is essential to successful sludge management and waste process design.The goal of this work was to establish a rapid routine analytical method to determine uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of up to 1000-fold higher total uranium concentrations (i.e., up to 30 wt% and more uranium) for samples to be taken during the upcoming sludge characterization campaign and in future analyses for sludge handling and processing. This report describes the experiments and results obtained in developing the selective dissolution technique to determine uranium metal concentration in K Basin sludge. The work described in this report:• provides a technical underpinning of the validity of the selective dissolution method, including the influence of various sludge components, in five test series using simulated and genuine sludges of widely varying composition• establishes analytical parameters (concentrations, quantities, temperatures, and times) necessary to develop the selective dissolution sludge-digestion analytical procedure.The uranium metal detection limit is estimated to be about 0.004 wt% based on the testing performed with actual sludge, thus meeting the goal detection limit of 0.03 wt%. The detection limit largely is established by the trace residual uranium carryover that exists in the leached and rinsed sludge. In the final series of method validation tests using actual K Basin sludge, spiked with uranium metal at concentrations ranging from 0.025 to 1.29 wt%, uranium metal recoveries averaged 99.4% with a standard deviation of 3.3%.

“…The UO 2 has been kept under water in its original closed jars since preparation until needed for experiment. Characterization tests showed UO 2 to be nearly stoichiometric, of nominal 6-nm diameter intrinsic particle size, and containing larger agglomerates (Sinkov et al 2008). The densities of the pure UO 2 and UO 3 ·2H 2 O phases are 10.964 g/cm 3 and 5.00 g/cm 3 , respectively (Grenthe et al 2006).…”

Section: 3mentioning

confidence: 98%

“…Mixtures of uranium dioxide (UO 2 ) and metaschoepite (UO 3 ·2H 2 O) were prepared by oxidation of UO 2 in aqueous suspension with a pure oxygen gas sparge (Sinkov et al 2008). The target oxidation state distribution was 50±5 mole% U(IV), as UO 2 , and 50±5 mole% U(VI), as UO 3 ·2H 2 O. Attainment of that distribution was established by spectrophotometry using an internal procedure.…”

Section: 3mentioning

confidence: 99%

Mitigation of Hydrogen Gas Generation from the Reaction of Uranium Metal with Water in K Basin Sludge and Sludge Waste Forms

Sinkov¹,

Delegard²,

2011

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Sludge formed and reposing in the K East (KE) and K West (KW) fuel storage basins on the Hanford Site has been a subject of concern since extended storage of metallic fuel from the N Reactor began in the 1980s. Although the last of the fuel was removed from the basins in 2004 and transport of sludge from the KE Basin to the KW Basin was completed in 2007, sludge still remains in the KW Basin stored in engineered containers. Sludge will be recovered from the engineered containers into Sludge Transport and Storage Containers and stored at an interim storage location on the Central Plateau. The stored sludge ultimately will be processed for disposal to the Waste Isolation Pilot Plant (WIPP) as remote-handled transuranic waste. However, before the sludge can go to WIPP, it must be processed and packaged into WIPP-acceptable forms. The presence of uranium metal fuel particles in the sludge poses problems to these operations because uranium metal reacts with water to form flammable hydrogen gas.Activities surrounding the storage, transport, treatment, and disposition of the K Basin sludge are currently administered under the Sludge Treatment Project (STP) managed by CH2M Hill Plateau Remediation Company (CHPRC) operating under contract to the U.S. Department of Energy. Investigative studies in this report were conducted by Pacific Northwest National Laboratory (PNNL) under contract to CHPRC for the STP.The reaction of uranium metal with water to form diatomic hydrogen gas (H 2 ) and uranium dioxide or uraninite (UO 2 ) proceeds by way of uranium hydride (UH 3 ). Mechanistic studies show that hydrogen radicals (H) and UH 3 serve as intermediates in the reaction of uranium metal with water to produce H 2 and UO 2 . Because H 2 is flammable, its release into the gas phase above K Basin sludge during sludge storage, processing, immobilization, shipment, and disposal is a concern to operational safety. PNNL conducted literature reviews and laboratory studies to identify methods that could be used to decrease the rate of H 2 gas generation from uranium metal corrosion in water present in sludge (Sinkov et al. 2010). Four means to decrease the H 2 evolution rate were identified for further experimental consideration:1. decreased temperature 2. reactant isolation (separation of the uranium metal from the water) 3. corrosion inhibition 4. hydrogen scavenging.The effect of decreased temperature could be inferred based on the known published rates of uranium metal reaction with water (Delegard and Schmidt 2009). Of the remaining three approaches, hydrogen scavenging appeared to be the most promising for laboratory testing, but experimental effort into reactant isolation and corrosion inhibition was also conducted and reported.Laboratory testing identified nitrate, NO 3 -, and nitrite, NO 2 -(added as their sodium salts) to be the most promising agents to minimize H 2 generation in prospective K Basin sludge operations because of their demonstrated high-attenuation factors, decreasing hydrogen generation rates hundreds-to thousan...