16 research outputs found
Separating the Minor Actinides Through Advances in Selective Coordination Chemistry
This report describes work conducted at the Pacific Northwest National Laboratory (PNNL) in Fiscal Year (FY) 2012 under the auspices of the Sigma Team for Minor Actinide Separation, funded by the U.S. Department of Energy Office of Nuclear Energy. Researchers at PNNL and Argonne National Laboratory (ANL) are investigating a simplified solvent extraction system for providing a single-step process to separate the minor actinide elements from acidic high-level liquid waste (HLW), including separating the minor actinides from the lanthanide fission products
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Separating the Minor Actinides Through Advances in Selective Coordination Chemistry
This report describes work conducted at the Pacific Northwest National Laboratory (PNNL) in Fiscal Year (FY) 2012 under the auspices of the Sigma Team for Minor Actinide Separation, funded by the U.S. Department of Energy Office of Nuclear Energy. Researchers at PNNL and Argonne National Laboratory (ANL) are investigating a simplified solvent extraction system for providing a single-step process to separate the minor actinide elements from acidic high-level liquid waste (HLW), including separating the minor actinides from the lanthanide fission products
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Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests
Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007)
Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests
Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007)
Distribution of Fission Products into Tributyl Phosphate under Applied Nuclear Fuel Recycling Conditions
Tributyl
phosphate (TBP) is an important industrial extractant
used in the Plutonium Uranium Redox Extraction (PUREX) process for
recovering uranium and plutonium from used nuclear fuel. Distribution
data have been assessed for a variety of fission and corrosion product
trace metals at varying uranium concentrations under representative
PUREX extraction (3 M HNO<sub>3</sub>) and stripping (0.1 M HNO<sub>3</sub>) conditions. As might have been anticipated, the extraction
of most trace metals was found to decrease or remain constant with
increasing uranium concentration. In contrast, the extraction of some
low valence transition metals was found to increase with increasing
uranium concentration. The increase in extraction of low valence transition
metals may be related to TBP forming reverse micelles instead of recovering
uranium as a classical UO<sub>2</sub>(NO<sub>3</sub>)<sub>2</sub>(TBP)<sub>2</sub> coordination complex. The low valence transition metals may
be being recovered into the cores of the reverse micelles. Also unanticipated
was the lack of impact the TBP degradation product, dibutyl phosphate
(DBP), had on the recovery of metals in batch distribution studies.
This is possibly related to the batch contacts completed in these
experiments not adequately recreating the multistage aspects of industrial-scale
uranium extraction done using mixer settlers or centrifugal contactors
Hexavalent Americium Recovery Using Copper(III) Periodate
Separation
of americium
from the lanthanides is considered one of the most difficult separation
steps in closing the nuclear fuel cycle. One approach to this separation
could involve oxidizing americium to the hexavalent state to form
a linear dioxo cation while the lanthanides remain as trivalent ions.
This work considers aqueous soluble Cu<sup>3+</sup> periodate as an
oxidant under molar nitric acid conditions to separate hexavalent
Am with diamyl amylphosphonate (DAAP) in <i>n</i>-dodecane.
Initial studies assessed the kinetics of Cu<sup>3+</sup> periodate
autoreduction in acidic media to aid in development of the solvent
extraction system. Following characterization of the Cu<sup>3+</sup> periodate oxidant, solvent extraction studies optimized the recovery
of Am from varied nitric acid media and in the presence of other fission
product, or fission product surrogate, species. Short aqueous/organic
contact times encouraged successful recovery of Am (distribution values
as high as 2) from nitric acid media in the absence of redox active
fission products. In the presence of a post-plutonium uranium redox
extraction (post-PUREX) simulant aqueous feed, precipitation of tetravalent
species (Ce, Ru, Zr) occurred and the distribution values of <sup>241</sup>Am were suppressed, suggesting some oxidizing capacity of
the Cu<sup>3+</sup> periodate is significantly consumed by other redox
active metals in the simulant. The manuscript demonstrates Cu<sup>3+</sup> periodate as a potentially viable oxidant for Am oxidation
and recovery and notes the consumption of oxidizing capacity observed
in the presence of the post-PUREX simulant feed will need to be addressed
for any approach seeking to oxidize Am for separations relevant to
the nuclear fuel cycle
Removing Phosphate from Hanford High-Phosphate Tank Wastes: FY 2010 Results
The U.S. Department of Energy (DOE) is responsible for environmental remediation at the Hanford Site in Washington State, a former nuclear weapons production site. Retrieving, processing, immobilizing, and disposing of the 2.2 × 105 m3 of radioactive wastes stored in the Hanford underground storage tanks dominates the overall environmental remediation effort at Hanford. The cornerstone of the tank waste remediation effort is the Hanford Tank Waste Treatment and Immobilization Plant (WTP). As currently designed, the capability of the WTP to treat and immobilize the Hanford tank wastes in the expected lifetime of the plant is questionable. For this reason, DOE has been pursuing supplemental treatment options for selected wastes. If implemented, these supplemental treatments will route certain waste components to processing and disposition pathways outside of WTP and thus will accelerate the overall Hanford tank waste remediation mission
Sigma Team for Minor Actinide Separation: PNNL FY 2011 Status Report
This report summarizes work conducted in FY 2011 at PNNL to investigate new methods of separating the minor actinide elements (Am and Cm) from the trivalent lanthanide elements, and separation of Am from Cm. For the former, work focused on a solvent extraction system combining an acidic extractant (HDEHP) with a neutral extractant (CMPO) to form a hybrid solvent extraction system referred to as TRUSPEAK (combining the TRUEX and TALSPEAK processes). For the latter, ligands that strongly bing uranyl ion were investigated for stabilizing corresponding americyl ion
Separating and Stabilizing Phosphate from High-Level Radioactive Waste: Process Development and Spectroscopic Monitoring
Removing phosphate from alkaline high-level waste sludges
at the
Department of Energy’s Hanford Site in Washington State is
necessary to increase the waste loading in the borosilicate glass
waste form that will be used to immobilize the highly radioactive
fraction of these wastes. We are developing a process which first
leaches phosphate from the high-level waste solids with aqueous sodium
hydroxide, and then isolates the phosphate by precipitation with calcium
oxide. Tests with actual tank waste confirmed that this process is
an effective method of phosphate removal from the sludge and offers
an additional option for managing the phosphorus in the Hanford tank
waste solids. The presence of vibrationally active species, such as
nitrate and phosphate ions, in the tank waste processing streams makes
the phosphate removal process an ideal candidate for monitoring by
Raman or infrared spectroscopic means. As a proof-of-principle demonstration,
Raman and Fourier transform infrared (FTIR) spectra were acquired
for all phases during a test of the process with actual tank waste.
Quantitative determination of phosphate, nitrate, and sulfate in the
liquid phases was achieved by Raman spectroscopy, demonstrating the
applicability of Raman spectroscopy for the monitoring of these species
in the tank waste process streams