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Plutonium dioxide dissolution in glass

Description: In the aftermath of the Cold War, the U.S. Department of Energy`s (DOE) Office of Fissile Materials Disposition (OFMD) is charged with providing technical support for evaluation of disposition options for excess fissile materials manufactured for the nation`s defense. One option being considered for the disposition of excess plutonium (Pu) is immobilization by vitrification. The vitrification option entails immobilizing Pu in a host glass and waste package that are criticality-safe (immune to nuclear criticality), proliferation-resistant, and environmentally acceptable for long-term storage or disposal. To prove the technical and economic feasibility of candidate vitrification options it is necessary to demonstrate that PuO{sub 2} feedstock can be dissolved in glass in sufficient quantity. The OFMD immobilization program has set a Pu solubility goal of 10 wt% in glass. The life cycle cost of the vitrification options are strongly influenced by the rate at which PUO{sub 2} dissolves in glass. The total number of process lines needed for vitrification of 50 t of Pu in 10 years is directly dependent upon the time required for Pu dissolution in glass. The objective of this joint Pacific Northwest National Laboratory (PNNL) - Savannah River Technology Center (SRTC) study was to demonstrate a high Pu solubility in glass and to identify on a rough scale the time required for Pu dissolution in the glass. This study was conducted using a lanthanide borosilicate (LaBS) glass composition designed at the SRTC for the vitrification of actinides.
Date: September 1, 1996
Creator: Vienna, J.D.; Alexander, D.L. & Li, Hong
Partner: UNT Libraries Government Documents Department

Drying Results of K-Basin Damaged/Corroded SNF Internal Sludge and Surface Coating

Description: Experiments were performed using a thermogravimetric analysis (TGA) system by Pacific Northwest National Laboratory (PNNL)to study the drying behavior of the K-Basin spent nuclear fuel (SNF) internal sludge and two different surface coatings of SNF elements. These measurements were conducted in support of the safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor fuel stored at K-Basin to an interim storage facility. These limited experiments on the corrosion products of K-Basin SNF material were part of the broad studies performed to ascertain the bounding pressurization of the Multi-Canister Overpack (MCO). Seven SNF internal sludge samples taken from different damage regions of three damaged/corroded outer K-Basin SNF elements were dried. Additionally, two surface coating samples taken from two SNF elements stored at K-West were tested. All the tests were performed in a vacuum atmosphere with the same temperature ramp rate of about 0.4 C/ min. Each TGA test sample was weighed before and after the test on a balance located in the Shielded Analytical Laboratory hot cell. The test samples were vacuum dried in the TGA system for about 24 hours prior to heating them at the rate of 0.4 C/min. The observations from the weight change data are summarized below.
Date: September 21, 2000
Creator: Abrefah, J.; Alexander, D.L. & Marschman, S.C.
Partner: UNT Libraries Government Documents Department

Defect Structure and Evolution in Silicon Carbide Irradiated to 1 dpa-SiC at 1100 C

Description: Transmission electron microscopy (TEM), swelling measurements, isochronal annealing, and thermal diffusivity testing were used to characterize the effects of radiation damage in SiC. Together, these techniques provided a comprehensive set of tools for observing and characterizing the structure and evolution of radiation-induced defects in SiC as a function of irradiation temperature and dose. In this study, two types of dense, crystalline, monolithic SiC were subjected to irradiation doses up to 1 dpa-SiC at a temperature of 1100 C, as well as post-irradiation annealing up to 1500 C. The microscopic defect structures observed by TEM were correlated to changes in the macroscopic dimensions, thermal diffusivity and thermal conductivity. The results demonstrated the value of using ultrapure {beta}SiC as an effective reference material to characterize the nature of expected radiation damage in other, more complex, SiC-based materials such as SiC/SiC composites.
Date: May 13, 2002
Creator: Senor, D.J.; Youngblood, G.E.; Greenwood, L.R.; Archer, D.V.; Alexander, D.L.; Chen, M.C. et al.
Partner: UNT Libraries Government Documents Department