Ion Exchange in High-Level Nuclear Waste Encapsulation: Atomistic Modeling of Equilibrium Coefficients and Isotherms

1999 ◽  
Vol 556 ◽  
Author(s):  
T. Kletskova ◽  
K. Czerwinski ◽  
E. Gelbard ◽  
S. Yip
Keyword(s):  
Ion Exchange ◽  
Interatomic Potential ◽  
Release Mechanism ◽  
Aqueous Environment ◽  
Atomistic Modeling ◽  
Modeling Approach ◽  
Simulation Techniques ◽  
High Level Nuclear Waste ◽  
Effects Of Radiation ◽  
High Level

AbstractA computational modeling approach, based on molecular dynamics and related atomistic simulation techniques, is formulated to analyze the confinement of radionuclides in prepared nuclear wasteform for long-term disposal. While the intent is to exploit the ability of simulation to provide a unified treatment of chemical as well as kinetic processes at the molecular level, it is recognized that the problem is too complex to be studied in its entirety at the outset. To demonstrate the feasibility and effectiveness of the atomistic modeling approach, a first step is to determine an equilibrium isotherm for cation exchange, on the premise that a predominant release mechanism is likely to involve ion exchange. As a prototypical application, Cs-Na exchange on sodalite is studied using explicit interatomic potential models. Initial results for the equilibrium Cs concentration in dehydrated sodalite suggest that the effects of frame relaxation need to be taken into account. The present study will be extended to treat an aqueous environment for the loaded wasteform, as well as the effects of radiation damage to the zeolitic framework.

Nonideality Effects on the Ion Exchange Behavior of the Zeolite Mineral Clinoptilolite

1990 ◽  
Vol 212 ◽  
Author(s):  
Roberto T. Pabalan
Keyword(s):  
Ion Exchange ◽  
Solid Phase ◽  
Solution Concentration ◽  
Nuclear Waste Repository ◽  
Thermodynamic Models ◽  
Diagenetic Alteration ◽  
Energy Models ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Induced Changes

ABSTRACTThe presence of laterally-extensive zones of zeolitized tuff underlying the proposed high-level nuclear waste repository at Yucca Mt., Nevada, has focused attention on the potential role of zeolite minerals, particularly clinoptilolite, in sorbing radionuclides and thereby retarding their migration. Ion exchange between zeolites and aqueous solutions depends on factors including compositions of the aqueous and zeolite phases and solution concentration. In addition, the thermodynamic stability of zeolite minerals and their susceptibility to diagenetic alteration also depend on aqueous and solid phase compositions. Therefore, spatial variations in zeolite compositions which have been observed at Yucca Mt., as well as natural or repository-induced changes in groundwater chemistry, may result in variations in the effectiveness of the zeolite minerals as retardation agents.Ion exchange experiments were conducted to obtain isotherm data and to evaluate the use of thermodynamic models in describing and predicting the solid solution and ion exchange properties of clinoptilolite. The experimental data were interpreted using excess Gibbs energy models for the aqueous solution and zeolite phases to account for nonideality in the system. The results indicate that the thermodynamic models allow predictions of clinoptilolite ion exchange behavior at ionic strengths and relative concentrations for which data are absent, and provide a foundation for the interpretation of ion exchange equilibria in multicomponent geochemical systems.

Selective removal of cesium and strontium using porous frameworks from high level nuclear waste

2016 ◽  
Vol 52 (35) ◽  
pp. 5940-5942 ◽  
Author(s):  
Briana Aguila ◽  
Debasis Banerjee ◽  
Zimin Nie ◽  
Yongsoon Shin ◽  
Shengqian Ma ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Aqueous Solutions ◽  
Nuclear Waste ◽  
Selective Removal ◽  
High Level Nuclear Waste ◽  
Presence And Absence ◽  
High Level ◽  
Exchange Properties ◽  
Competing Ions

A water stable MOF, MIL-101-SO3H, shows excellent Cs+ and Sr2+ ion exchange properties in aqueous solutions in the presence and absence of competing ions.

Initial Demonstration of the Vitrification of High-Level Nuclear Waste Sludge Containing an Organic Cs-Loaded Ion-Exchange Resin

1992 ◽  
Vol 294 ◽  
Author(s):  
N. E. Bibler ◽  
J. P. Bibler ◽  
M. K. Andrews ◽  
C. M. Jantzen
Keyword(s):  
Ion Exchange ◽  
Nuclear Waste ◽  
Borosilicate Glass ◽  
Exchange Resin ◽  
Ion Exchange Resin ◽  
Waste Sludge ◽  
New Process ◽  
The Usa ◽  
High Level Nuclear Waste ◽  
High Level

ABSTRACTWhen immobilizing into borosilicate glass the radionuclides in the caustic high-level radioactive wastes stored in the USA, the soluble fission product Cs-137 has to be removed from supernates of the wastes. In the current processes zeolites or an organic precipitant will be used to remove the Cs. These are then treated further and mixed with the radioactive sludges and vitrified into a borosilicate glass. This paper describes the vitrification of a mixture resulting from using a new process to remove Cs from the caustic supernate. A resorcinol based organic ion exchange resin is used. This resin was then mixed with sludge and frit and vitrified. Using an organic ion exchange resin rather than zeolite or the organic precipitant has certain advantages. For example, use of the zeolite increases the amount of glass to be made and use of the organic precipitant produces benzene as a secondary waste stream. Results in the paper indicate that a mixture of the resin, sludge and frit can be successfully vitrified in a joule-heated, slurry fed melter. However, when resin is present in the feed, the glass becomes less durable due to the increased amount of Fe(II) caused by reduction of Fe(III) in the melt. Based on the durabilities of other waste glasses, this glass is still suitable as a canistered wasteform.

Mcc Corrosion Tests at Reference Testing Conditions for A27 Cast steel in Hanford Ground Water

1986 ◽  
Vol 84 ◽  
Author(s):  
M.D. Merz ◽  
F. Gerber ◽  
R. Wang
Keyword(s):  
Pacific Northwest ◽  
Pressure Vessel ◽  
Static Pressure ◽  
Cast Steel ◽  
Statistical Treatment ◽  
Testing Conditions ◽  
Corrosion Tests ◽  
Corrosion Rates ◽  
High Level Nuclear Waste ◽  
High Level

AbstractThe Materials Characterization Center (MCC) at Pacific Northwest Lab- oratory is performing three kinds of corrosion tests for the Basalt Waste Isolation Project (BWIP) to establish the interlaboratory reproducibility and uncertainty of corrosion rates of container materials for high-level nuclear waste. The three types of corrosion tests were selected to address two distinct conditions that are expected in a repository constructed in basalt. An air/steam test is designed to address corrosion during the operational period and static pressure vessel and flowby tests are designed to address corrosion under conditions that bound the condi ring the post-closure period of the repository.The results of tests at reference testing conditions, which were defined to facilitate interlaboratory comparison of data, are presented. Data are reported for the BWIP/MCC-105.5 Air/Steam Test, BWIP/MCC-105.1 Static Pressure Vessel, and BWIP/MC-105.4 Flowby Test. In those cases where data are available from a second laboratory, a statistical analysis of interlaboratory results is reported and expected confidence intervals for mean corrosion rates are given. Other statistical treatment of data include analyses of the effects of vessel-to-vessel variations, test capsule variations for the flowby test, and oven-to-oven variations for air/steam tests.

A Perspective on Modeling Materials in Extreme Environments: Oxidation of Ultrahigh-Temperature Ceramics

MRS Bulletin ◽  
2006 ◽  
Vol 31 (5) ◽  
pp. 410-418 ◽  
Author(s):  
Angelo Bongiorno ◽  
Clemens J. Först ◽  
Rajiv K. Kalia ◽  
Ju Li ◽  
Jochen Marschall ◽  
...  
Keyword(s):  
Modeling And Simulation ◽  
Atomistic Simulation ◽  
Extreme Environments ◽  
Protective Layer ◽  
Ceramic Composites ◽  
Atomistic Modeling ◽  
Temperature Oxidation ◽  
Simulation Techniques ◽  
Oxidation Resistant ◽  
Ultrahigh Temperature

AbstractThe broader context of this discussion, based on a workshop where materials technologists and computational scientists engaged in a dialogue, is an awareness that modeling and simulation techniques and computational capabilities may have matured sufficiently to provide heretofore unavailable insights into the complex microstructural evolution of materials in extreme environments.As an example, this article examines the study of ultrahigh-temperature oxidation-resistant ceramics, through the combination of atomistic simulation and selected experiments.We describe a strategy to investigate oxygen transport through a multi-oxide scale—the protective layer of ultrahigh-temperature ceramic composites ZrB2-SiC and HfB2-SiC—by combining first-principles and atomistic modeling and simulation with selected experiments.

Conductivity Behavior of Salt Deposits on the Surface of Engineered Barrier Materials for the Potential High-Level Nuclear Waste Repository at Yucca Mountain, Nevada

2004 ◽  
Vol 824 ◽  
Author(s):  
Lietai Yang ◽  
Miriam R. Juckett ◽  
Roberto T. Pabalan
Keyword(s):  
Electrical Conductance ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Conductivity Measurements ◽  
Barrier Materials ◽  
Salt Mixtures ◽  
High Level Nuclear Waste ◽  
Waste Repository ◽  
High Level ◽  
Conductivity Behavior

AbstractThe electrical conductance or conductivity of three salt mixtures, Na-K-Cl-NO3, Ca-K-Cl and Ca-Na-Cl, were measured at 25, 50 and 70°C [77, 122, and 158 °F] as a function of relative humidity (RH). Mutual deliquescence and efflorescence RH (MDRH and MERH) values were determined based on the conductivity measurements. It was found that the conductivity of the three salt mixtures started to increase at RH values that are approximately 40 % of their MDRH and increased by 1to 2 orders of magnitude just before reaching the MDRH. At the MDRH, a significant increase in conductivity was observed. The MDRH and MERH for the Ca-K-Cl and Ca-Na-Cl mixtures were found to be approximately 15 % in the temperature range of 50 to 70 °C [122 to 158 °F]. The MDRH and MERH for the Na-K-Cl-NO3system were found to be approximately 54 % at 50 °C [122 °F] and decreased significantly with an increase in temperature.

Transmutation of high-level nuclear waste by means of accelerator driven system

2013 ◽  
Vol 3 (1) ◽  
pp. 60-69 ◽  
Author(s):  
Hamid Aït Abderrahim ◽  
Didier De Bruyn ◽  
Gert Van den Eynde ◽  
Sidney Michiels
Keyword(s):  
Nuclear Waste ◽  
Driven System ◽  
High Level Nuclear Waste ◽  
Accelerator Driven System ◽  
High Level
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