Impact of Irradiation on Solvent used in SRS Waste Treatment Processes -9122

ABSTRACT Savannah River Site (SRS) will use a Caustic Side Solvent Extraction (CSSX) process to selectively remove radioactive Cs-137 from the caustic High Level Waste (HLW) salt solutions stored in the large carbon steel waste tanks in the SRS Tank Farm. This HLW resulted from several decades of operations at SRS to produce nuclear materials for the United States Government. The removed Cs-137 will be sent to the Defense Waste Processing Facility (DWPF) where it will be immobilized along with the HLW sludges from the SRS Tank Farm into a borosilicate glass that will be put into permanent disposal. Currently the CSSX process is operating on an interim basis in the Modular Caustic Side Solvent Extraction Unit (MCU) facility. Eventually the process will occur in the full scale Salt Waste Processing Facility (SWPF) currently being built. The organic solvent developed for the process is primarily a mixture of the Isopar ® L (a blend of C 10 -C 12 branched alkanes such as dodecane) and an alkyl aryl polyether added as a Modifier (commonly called Cs-7SB) to enhance the solubility of the extractant which is a calixarene-crown ether. The solvent also includes trioctylamine to mitigate the adverse impact of lipophilic agents on the stripping of the cesium into nitric acid. Since the mixture is primarily organic hydrocarbons, it is expected that radiolysis of the mixture with gamma rays and beta particles from the Cs-137 will produce the flammable gas H 2 and also eventually degrade the solvent. For example, much research has been performed on the radiolysis of the organic solvent used in the tributylphosphate (TPB) extraction process (PUREX process) that has been used at SRS and in many other countries for several decades to separate U and Pu from radioactive U-235 fission products such as Cs-137. [1] The purpose of this study was to investigate the radiolysis of the organic solvent for the CSSX process. Researchers at Savannah River National Laboratory (SRNL) irradiated samples of solvent with Co-60 gamma rays. Prior to the irradiation, the solvent was contacted with the aqueous solutions that will be used in the MCU and SWPF facilities. These were the aqueous caustic salt feed, the scrub solution, and wash water. The rates of radiolytic H 2 production were measured both by determining the composition of the gases produced and by measuring pressures produced during radiolysis. The irradiated solvents were then analyzed by various analytical techniques to assess how much of the Isopar ® L, the Modifier, and the extractant had decomposed

Performance of a Buried Radioactive High Level Waste Glass After 24 Years

A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet

EVALUATION OF IMPURITY EXTREMES IN A PLUTONIUM-LOADED BOROSILICATE GLASS

A vitrification technology utilizing a lanthanide borosilicate (LaBS) glass appears to be a viable option for the disposition of excess weapons-useable plutonium that is not suitable for processing into mixed oxide (MOX) fuel. A significant effort to develop a glass formulation and vitrification process to immobilize plutonium was completed in the mid-1990s. The LaBS glass formulation was found to be capable of immobilizing in excess of 10 wt % Pu and to be tolerant of a range of impurities. To confirm the results of previous testing with surrogate Pu feeds containing impurities, four glass compositions were selected for fabrication with actual plutonium oxide and impurities. The four compositions represented extremes in impurity type and concentration. The homogeneity and durability of these four compositions were measured. The homogeneity of the glasses was evaluated using x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS). The XRD results indicated that the glasses were amorphous with no evidence of crystalline species in the glass. The SEM/EDS analyses did show the presence of some undissolved PuO{sub 2} material. The EDS spectra indicated that some of the PuO{sub 2} crystals also contained hafnium oxide. The SEM/EDS analyses showed that there were no heterogeneities in the glass due to the feed impurities. The durability of the glasses was measured using the Product Consistency Test (PCT). The PCT results indicated that the durability of Pu impurity glasses was comparable with Pu glasses without impurities and significantly more durable than the Environmental Assessment (EA) glass used as the benchmark for repository disposition of high-level waste (HLW) glasses

The Product Consistency Test How and Why It Was Developed

The Product Consistency Test (PCT), American Society for Testing Materials (ASTM) Standard C1285, is currently used world wide for testing glass and glass-ceramic waste forms for high level waste (HLW), low level waste (LLW), and hazardous wastes. Development of the PCT was initiated in 1986 because HLW glass waste forms required extensive characterization before actual production began and required continued characterization during production ({ge}25 years). Non-radioactive startup was in 1994 and radioactive startup was in 1996. The PCT underwent extensive development from 1986-1994 and became an ASTM consensus standard in 1994. During the extensive laboratory testing and inter- and intra-laboratory round robins using non-radioactive and radioactive glasses, the PCT was shown to be very reproducible, to yield reliable results rapidly, to distinguish between glasses of different durability and homogeneity, and to easily be performed in shielded cell facilities with radioactive samples. In 1997, the scope was broadened to include hazardous and mixed (radioactive and hazardous) waste glasses. In 2002, the scope was broadened to include glass-ceramic waste forms which are currently being recommended for second generation nuclear wastes yet to be generated in the nuclear renaissance. Since the PCT has proven useful for glass-ceramics with up to 75% ceramic component and has been used to evaluate Pu ceramic waste forms, the use of this test for other ceramic/mineral waste forms such as geopolymers, hydroceramics, and fluidized bed steam reformer mineralized product is under investigation

Role of groundwater oxidation potential and radiolysis on waste glass performance in crystalline repository environments

Laboratory experiments have shown that groundwater conditions in a Stripa granite repository will be as reducing as those in a basalt repository. The final oxidation potential (Eh) at 70/sup 0/C for Stripa groundwater deaerated and equilibrated with crystalline granite was -0.45V. In contrast, the oxidation potential at 60/sup 0/C for Grande Ronde groundwater equilibrated with basalt was -0.40V. The reducing groundwater conditions were found to slightly decrease the time-dependent release of soluble components from the waste glass. Spectrophotometric analysis of the equilibrated groundwaters indicated the presence of Fe/sup 2 +/ confirming that the Fe/sup 2 +//Fe/sup 3 +/ couple is controlling the oxidation potential. It was also shown that in the alkaline pH regime of these groundwaters the iron species are primarily associated with x-ray amorphous precipitates in the groundwater. Gamma radiolysis in the absence of waste glass and in the absence of oxygen further reduces the oxidation potential of both granitic and basaltic groundwaters. The effect is more pronounced in the basaltic groundwater. The mechanism for this decrease is under investigation but appears related to the reactive amorphous precipitate. The results of these tests suggest that H/sub 2/ may not escape from the repository system as postulated and that radiolysis may not cause the groundwaters to become oxidizing in a crystalline repository when abundant Fe/sup 2 +/ species are present. 23 refs., 3 figs., 3 tabs

Multimodal separation of alkali, alkaline earth, transition, post-transition, lanthanide, and actinide metal cations in waste sludge

An ion chromatographic method, which separates 36 different cations in a single chromatographic run, was developed to separate and analyze trace radionuclides present on high level radioactive waste samples. The method employs linear and step gradients and isocratic elution using four different eluents in six different eluent phases. The separation takes 45 minutes and has detection limits ranging from 0.1 ppM to 5.0 ppM, when using spectrophotometric detection for nonradioactive cations, depending on the sample matrix. The detection limits and relative standard deviation of the data are dependent upon the element and sample matrix. This method can be reliably performed in the laboratory if properly prepared samples are used. This study describes the applications, limitations, interferences, precision, and accuracy of this method. Using this method, trace radionuclides, which are present in concentrations of only a few hundred disintegrations per minute per milliliter, can be separated and then analyzed by using liquid scintillation counting analysis and inductively coupled plasma mass spectrometry (ICP-MS). This paper will first describe the chromatographic separation as it was developed and applied to the analysis of aqueous samples with low ppM levels of nonradioactive cations. Next, the application of this method to the separation and analysis of high level waste tank samples will be discussed