Elemental dissolution study of Pu-bearing borosilicate glasses

Wellman, Dawn M.; Icenhower, Jonathan P.; Weber, William J.

doi:10.1016/j.jnucmat.2004.10.166

Cited by 53 publications

(37 citation statements)

References 35 publications

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“…For example, tests were conducted with a borosilicate HLW glass to study the effect of self-radiation (Wellman et al, 2005). Those tests were conducted with borosilicate glass samples that had been made about 20 years previously with different mixtures of Pu-239 and Pu-238 (1 mass % total PuO 2 ) and had achieved different accumulated doses up to 2.6 × 10 18 alpha decays/g when the measurements were made.…”

Section: Ivb6 Radiation Damagementioning

confidence: 99%

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Ebert¹

2006

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Section: Ivb6 Radiation Damagementioning

confidence: 99%

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Ebert¹

2006

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“…The SPFT method has been described in detail elsewhere (McGrail et al 1997b;Wellman et al 2006a;Wellman et al 2005); therefore, only a brief description will be provided here and interested readers should consult the noted references for additional information. In general, the SPFT system ( Figure 3.2) consists of a programmable pump (Kloehn; model 50300) that transports solutions from an influent reservoir via Teflon lines.…”

Section: Single-pass Flow-through Test Methodsmentioning

confidence: 99%

DNA Profile Analysis and Interpretation

2014

DNA Analysis for Missing Person Identification in Mass Fatalities

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SummaryThe Hanford Site is a former nuclear defense production facility. A groundwater plume containing uranium, originating from a combination of purposeful discharges of wastewater to cribs, trenches, and ponds, along with some accidental leaks and spills related to nuclear fuel fabrication activities, has persisted beneath the Hanford Site 300 Area for many years. Despite the cessation of uranium releases and the removal of shallow vadose zone source materials, the remedial action objective to lower the concentration of groundwater uranium to the U. S. Environmental Protection Agency maximum contaminant level concentration of 30 µg/L has not been achieved within the anticipated 10-year time period. Some unknown amount of contamination remains in the vadose zone beneath the lower extent of the excavation activities. Additional contamination also may remain beneath buildings and facilities in the southern portion of the 300 Area, which has not been decontaminated and decommissioned. The use of polyphosphate technology for source treatment in the vadose zone and capillary fringe is expected to accelerate the natural attenuation of uranium to thermodynamically stable uranium-phosphate minerals. This effort will complement the current 300 Area treatability test being conducted within the saturated zone (e.g., 300 Area aquifer) for in situ treatment of uranium-contaminated groundwater.Polyphosphate technology has been demonstrated for in situ precipitation of phosphate phases to control the long-term fate of uranium. A critical component of the development and testing is detailed evaluation to determine if polyphosphate technology could be modified for infiltration from ground surface or some depth of excavation to stabilize source uranium phases. This report presents results from bench-scale treatability studies conducted under site-specific conditions to optimize the polyphosphate amendment for implementation of a field-scale technology demonstration to stabilize uranium within the 300 Area vadose zone and capillary fringe of the Hanford Site. Documented in this report are data related to 1) the retardation of polyphosphate as a function of water content and pore water velocity, 2) the reaction between uranium-bearing solid phases and aqueous polyphosphate remediation technology as a function of polyphosphate composition and concentration, 3) the mechanism of autunite formation via the reaction of solid-phase calcite-bound uranium and aqueous polyphosphate remediation technology, 4) the transformation mechanism and reaction kinetics between uranyl-carbonate and -silicate minerals with the polyphosphate remedy under advective conditions, and 5) the extent and rate of uranium released and immobilized as a function of polyphosphate composition and the infiltration rate of the polyphosphate remedy. Kinetic rate law parameters were determined from single-pass flow-through experiments. Pressurized unsaturated flow tests were used to determine the effect of polyphosphate composition, concentration, and infiltration rate...

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“…This conclusion is confirmed by many experiments; for example, unaffected corrosion behavior has recently been found for a Pu-bearing borosilicate glass over a pH interval of 9 to 12 at 353 K to 361 K (80°C to 88°C). [70] Leach tests of French SON68 glass in silica-saturated solutions showed that ion-exchange rates are increased after irradiation whereas hydrolysis remained unchanged. [71] Similar effects were observed in irradiated phosphate glasses.…”

Section: Effects Of Self-irradiationmentioning

confidence: 99%

Glassy Wasteforms for Nuclear Waste Immobilization

Ojovan

Lee

2010

Metall Mater Trans A

246

124

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Glassy wasteforms currently being used for high-level radioactive waste (HLW) as well as for low-and intermediate-level radioactive waste (LILW) immobilization are discussed and their most important parameters are examined, along with a brief description of waste vitrification technology currently used worldwide. Recent developments in advanced nuclear wasteforms are described such as polyphase glass composite materials (GCMs) with higher versatility and waste loading. Aqueous performance of glassy materials is analyzed with a detailed analysis of the role of ion exchange and hydrolysis, and performance of irradiated glasses.

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Elemental dissolution study of Pu-bearing borosilicate glasses

Cited by 53 publications

References 35 publications

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

DNA Profile Analysis and Interpretation

Glassy Wasteforms for Nuclear Waste Immobilization

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