The Leaching Behavior of a Glass Waste Form – Part I: The Characteristics of Surface Layers

1984 ◽  
Vol 67 (3) ◽  
pp. 419-428 ◽  
Author(s):  
Takashi Murakami ◽  
Tsunetaka Banba
Keyword(s):  
Waste Form ◽  
Surface Layers ◽  
Leaching Behavior ◽  
Glass Waste ◽  
Form Part

The Leaching Behavior of a Glass Waste Form—Part III: The Mathematical Leaching Model

1987 ◽  
Vol 76 (1) ◽  
pp. 84-90 ◽  
Author(s):  
Tsunetaka Banba ◽  
Takashi Murakami ◽  
Hideo Kimura
Keyword(s):  
Waste Form ◽  
Leaching Behavior ◽  
Glass Waste ◽  
Form Part

The Leaching Behavior of a Glass Waste Form–Part II: The Leaching Mechanisms

1985 ◽  
Vol 70 (2) ◽  
pp. 243-248 ◽  
Author(s):  
Tsunetaka Banba ◽  
Takashi Murakami
Keyword(s):  
Waste Form ◽  
Leaching Behavior ◽  
Glass Waste ◽  
Form Part

Crystalline Phase in Surface Layers of Glass Waste Form

1983 ◽  
Vol 20 (7) ◽  
pp. 614-617
Author(s):  
Takashi MURAKAMI ◽  
Tsunetaka BANBA
Keyword(s):  
Crystalline Phase ◽  
Waste Form ◽  
Surface Layers ◽  
Glass Waste

Integrated Testing of the Srl-165 Glass Waste Form

1986 ◽  
Vol 84 ◽  
Author(s):  
D.L. Phinney ◽  
F.J. Ryerson ◽  
V.M. Oversby ◽  
W.A. Lanford ◽  
R.D. Aines ◽  
...  
Keyword(s):  
Mass Balance ◽  
High Surface ◽  
High Energy ◽  
Leach Solution ◽  
Solution Chemistry ◽  
Waste Form ◽  
Depth Of Penetration ◽  
Glass Waste ◽  
Integrated Testing ◽  
Radionuclide Release

AbstractIntegrated testing of the important components of a glass waste form waste package has been performed in order to gain a better understanding of the processes of radionuclide release and transport in the near field environment. Based upon an interpretation of the depth of penetration of hydrogen in reacted SRL-165 glass we have modeled the radionuclide release from the glass as a combined process of (1) the diffusive exchange of alkalis and boron in the glass for hydrogen species in the solution (D=10−16 cm2/s) and (2) surface dissolution. Surface dissolution controls the release of components not exchanged by diffusion and takes place at a rate of 1.5-3.0 μm/yr. Subsequent to release the radionuclides may remain in the leach solution, diffuse into the tuff, or precipitate as secondary phases. Precipitation is particularly important for plutonium and americium. Diffusive transport of radionuclides through the tuff takes place at an extremely slow rate, D=10−16 cm2/s. As such, the mass of radionuclides incorporated in the tuff by diffusion during the tests is inconsequential relative to that in the leach solution (with the exception of plutonium) and can be ignored in mass balance calculations. Mass balance calculations based upon the release of radionuclides by surface dissolution of the glass waste form are in good agreement with observed solution chemistry when allowances are made for a pulse of dissolution early in the tests. This pulse may be due to either the rapid dissolution of high-energy surface features early in the inLegrated tests, or an initially high surface dissolution rate that decreases with time as silica saturation is approached [1], or a combination of the two.

Materials Interactions Relating to Long-Term Geologic Disposal of Nuclear Waste Glass

1986 ◽  
Vol 84 ◽  
Author(s):  
Ned E. Bibler ◽  
Carol M. Jantzen
Keyword(s):  
Nuclear Waste ◽  
Review Paper ◽  
Waste Glass ◽  
Waste Form ◽  
Glass Waste ◽  
Waste Package ◽  
Geologic Disposal ◽  
Nuclear Waste Glass ◽  
Backfill Materials

AbstractIn the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs. anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made.

Wipp Waste Package Testing on Simulated DHLW: Emplacement

1984 ◽  
Vol 44 ◽  
Author(s):  
Martin A. Molecke
Keyword(s):  
Thermal Loading ◽  
Accelerated Aging ◽  
Waste Form ◽  
Future Application ◽  
High Level Waste ◽  
Glass Waste ◽  
Waste Package ◽  
High Level ◽  
Artificial Introduction ◽  
Materials Behavior

AbstractSeveral series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have recently been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests.These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs.These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplaced under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced tinder accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass.

Effect of surfaces on glass waste form leaching

1982 ◽  
Vol 49 (1-3) ◽  
pp. 379-388 ◽  
Author(s):  
C Pescatore ◽  
A.J Machiels
Keyword(s):  
Waste Form ◽  
Glass Waste

Microstructure and Chemistry of an Aluminophosphate Glass Waste Form under Electron Beam Irradiation

2003 ◽  
Vol 807 ◽  
Author(s):  
K. Sun ◽  
L. M. Wang ◽  
R. C. Ewing
Keyword(s):  
Electron Irradiation ◽  
Analytical Electron Microscopy ◽  
Iron Phosphate ◽  
Phosphate Glasses ◽  
Waste Form ◽  
Irradiation Damage ◽  
Glass Waste ◽  
Electron Dose ◽  
Aluminophosphate Glass ◽  
Iron Phosphate Glasses

ABSTRACTAn alkali-containing (mainly sodium) aluminophosphate glass waste form was studied by analytical electron microscopy. Bright-field imaging showed that small bubbles formed under the electron irradiation even at a low electron dose (8×1022e/m2). These bubbles grew with the increase of the electron dose and were finally released at the surface of the sample. At the same time alkali elements were also lost under irradiation. At an electron dose of about 2.2×1026e/m2 (6.6×1011Gy), all the bubbles were released and no bubbles were formed with further irradiation. The glass was finally transformed to a pure aluminophosphate glass. Further irradiation resulted in the phase separation between the Al-rich and the P-rich phases. The electron irradiation damage effects on the aluminophosphate glass are compared to those observed in iron phosphate glasses.

Long-Term Radioactivity Release from Solidified High-Level Waste - Part II Parametric Study of Waste form Properties, Temperature and Time

1982 ◽  
Vol 15 ◽  
Author(s):  
Friedrich K. Altenhein ◽  
Werner Lutze ◽  
Rodney C. Ewing
Keyword(s):  
Parametric Study ◽  
Computer Code ◽  
Salt Dome ◽  
Waste Form ◽  
Glass Block ◽  
High Level Waste ◽  
Surface Layers ◽  
History Of ◽  
High Level

The computer code QTERM has been used to calculate the total released activity from a single glass block when in contact with brine in a salt dome repository as a function of: (1) waste form properties, (2) leaching mechanisms, (3) retention or precipitation of specific radionuclides in surface layers, (4) thermal history of the repository and (5) decreasing activity as a function of time.

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