scholarly journals Forty years of durability assessment of nuclear waste glass by standard methods

2021 ◽  
Vol 5 (1) ◽  
Author(s):  
Clare L. Thorpe ◽  
James J. Neeway ◽  
Carolyn I. Pearce ◽  
Russell J. Hand ◽  
Adam J. Fisher ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Nuclear Waste ◽  
Glass Composition ◽  
Radioactive Waste Disposal ◽  
Experimental Conditions ◽  
Standard Methods ◽  
Glass Dissolution ◽  
Nuclear Waste Glass ◽  
Safety Case ◽  
Durability Assessment

AbstractStandard methods to assess the durability of vitrified radioactive waste were first developed in the 1980’s and, over the last 40 years, have evolved to yield a range of responses depending on experimental conditions and glass composition. Mechanistic understanding of glass dissolution has progressed in parallel, enhancing our interpretation of the data acquired. With the implementation of subsurface disposal for vitrified radioactive waste drawing closer, it is timely to review the available standard methodologies and reflect upon their relative advantages, limitations, and how the data obtained can be interpreted to support the post-closure safety case for radioactive waste disposal.

The effect of phosphate melt cooling rate on phase composition and leach resistance of final waste form

MRS Advances ◽  
2017 ◽  
Vol 3 (20) ◽  
pp. 1085-1091 ◽  
Author(s):  
Konstantin V. Martynov ◽  
Elena V. Zakharova ◽  
Sergey V. Stefanovsky ◽  
Boris F. Myasoedov
Keyword(s):  
Nuclear Waste ◽  
Glass Composition ◽  
Crystalline Material ◽  
Slow Cooling ◽  
Leach Rate ◽  
Glass Dissolution ◽  
Phosphate Phase ◽  
Partial Crystallization ◽  
Melt Cooling ◽  
Melt Solidification

ABSTRACTSlow cooling of phosphate melt at liquid nuclear waste solidification yields glass-crystalline material. Partial crystallization during melt solidification results in elemental partitioning among crystalline phase and glass: Al, Cr, Fe are concentrated in the crystalline phosphate phase while Ca, Ni, La, U enter predominantly in the residual glass. Glass dissolution rate and leach rate of La and U as rare earth and actinide surrogates depends strongly on the glass composition, for example reduction of Al2O3 content in the glass to ∼10-12 wt.% increases leachability by three orders of magnitude as compared to the glass with specified composition (∼18-22 wt.% Al2O3).

scholarly journals A tentative useful-for-all definition of “safety” in the context of radioactive waste disposal

2021 ◽  
Vol 1 ◽  
pp. 195-196
Author(s):  
Stephan Hotzel
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Waste Disposal ◽  
Common Ground ◽  
Radioactive Waste Disposal ◽  
National Programmes ◽  
Geological Repository ◽  
Safety Case ◽  
Definition Of

Abstract. Most, if not all, national programmes for radioactive waste management pledge their overall commitment to safety or – in the case of radioactive waste disposal – to long-term safety. Therefore, it may be somewhat surprising to find that the term “safety” is hardly defined in these programs. The same holds for some of the core international guidance literature on the deep geological repository (DGR) “safety case” concept. With respect to stakeholder concern over the safety of geological disposal, it seems, however, advisable to seek common ground in the understanding of the idea of “safety”. Hotzel and Schröder (2018) reviewed the most relevant international guidance literature for explicitly or implicitly provided definitions of “safety” in the context of radioactive waste disposal. Based on this study – and on the finding that a practical, useful-for-all definition of “safety” is not provided in the scanned literature – they developed a tentative dictionary-style definition of “safety” that is suitable for everyday use in the DGR context. In the current contribution I embed, expand and update the 2018 study at both ends: As an enhanced introduction to the 2018 study, I lay out a basic concept of “sound” glossary definitions, namely glossary definitions being both practical and correct (and what this means). The thesis is that sound glossary definitions can facilitate mutual understanding between different stakeholder groups. As an update to the actual proposal for the definition of “safety” from the Hotzel and Schröder (2018) paper, that was presented and discussed at the Waste Management Conference 2018, I review the latest international guidance literature and the stakeholder concerns raised at the 2018 conference in order to present a revised definition. As a seed of discussion, it may help to eventually expose possible mismatches in the base assumptions of safety experts and other stakeholders and thereby support meaningful communication.

The Interaction Between Nuclear Waste Glasses and Clay-II

1989 ◽  
Vol 176 ◽  
Author(s):  
P. Van Iseghem ◽  
K. Berghman ◽  
W. Timmermans
Keyword(s):  
Detection Limit ◽  
Corrosion Rate ◽  
Diffusion Process ◽  
Gamma Irradiation ◽  
Nuclear Waste ◽  
Experimental Conditions ◽  
Glass Dissolution ◽  
Glass Corrosion ◽  
Irradiation Field

ABSTRACTThe interaction between simulated HLW glasses and clay has been investigated, as part of a larger programme to evaluate the performance of vitrified HLW in repository conditions. Experimental conditions were carefully chosen to simulate repository conditions as close as possible. The glass corrosion seems to be controlled by a diffusion process, but there are indications of a (small) final corrosion rate. The leaching of 134Cs, 90Sr, 239Pu and 241Am is characterized by extremely small mobile concentrations leached (for 239pu and 241Am, below detection limit). The presence of a gamma irradiation field does not increase the glass dissolution. The production of radiolythic hydrogen is strongly reduced by the presence of clay.

A Kinetic Model of Nuclear Waste Glass Dissolution in Flowing Water Environments

1984 ◽  
Vol 44 ◽  
Author(s):  
Aaron Barkatt ◽  
Barbara C. Gibson ◽  
Marek Brandys
Keyword(s):  
Kinetic Model ◽  
Nuclear Waste ◽  
Flow Rate ◽  
Flowing Water ◽  
Flow Rates ◽  
Glass Dissolution ◽  
Nuclear Waste Glass ◽  
Simple Kinetic Model ◽  
Leachate Composition ◽  
Flow Experiments

AbstractA simple kinetic model for the description of the interaction of nuclear waste borosilicate glass with water has been developed. In the case of SRL TDS-131 glass leached in water at 70°C over a broad range of flow rates this model was found useful both in describing the evolution of leachate composition as a function of exposure time at a particular flow rate and in describing the dependence of the steady-state concentrations of the leached elements in solution on flow rate in a series of flow experiments.

Mineralogical aspects of cement in radioactive waste disposal

2001 ◽  
Vol 65 (5) ◽  
pp. 621-633 ◽  
Author(s):  
F. P. Glasser
Keyword(s):  
Radioactive Waste ◽  
Nuclear Waste ◽  
Radioactive Waste Disposal ◽  
Geological Time ◽  
Hydrous Oxides ◽  
Physical Chemical ◽  
Mineralogical Characteristics ◽  
Waste Immobilization ◽  
Nuclear Waste Immobilization

AbstractSome of the physical, chemical and mineralogical characteristics of Portland cement and related materials relevant to nuclear waste immobilization are defined. The ability to condition and maintain a high aqueous pH is undoubtedly the most important factor: it precipitates many species as hydrous oxides or hydroxides. However, in the longer term, many species – cationic as well as anionic – react with one or more cement components forming solubility-limiting phases. Progress on characterization of these phases is outlined. Many of the host phases have natural equivalents and this gives comfort in respect of their likely persistence over geological time. The emerging picture of immobilization in cement suggests that cement compositions can be tailored in terms of pH, Eh and internal chemistry so as to maximize immobilization potential. Nickel, uranium and chromium and chloride are used as examples.

scholarly journals Towards best possible safety – current regulatory research for the German site selection process for high-level radioactive waste disposal

2020 ◽  
Vol 54 ◽  
pp. 157-163
Author(s):  
Axel Liebscher ◽  
Christoph Borkel ◽  
Michael Jendras ◽  
Ute Maurer-Rurack ◽  
Carsten Rücker
Keyword(s):  
Radioactive Waste ◽  
Nuclear Waste ◽  
Waste Disposal ◽  
Rock Salt ◽  
Site Selection ◽  
State Of The Art ◽  
Selection Process ◽  
Fault Zones ◽  
Radioactive Waste Disposal ◽  
Methodological Approaches

Abstract. The Federal Office for the Safety of Nuclear Waste Management (BASE – Bundesamt für die Sicherheit der nuklearen Entsorgung) is the German federal regulatory authority for radioactive waste disposal. It supervises the German site selection process and is responsible for the accompanying public participation. Task related research is an integral part of BASE's activities. The projects MessEr and übErStand compiled the state-of-the-art science and technology regarding surface based exploration methods suitable for addressing the criteria and requirements specified in the German Site Selection Act. The results support BASE to review and define the surface-based exploration programs to be executed by the national implementer BGE (Bundesgesellschaft für Endlagerung mbH). To support BASE in reviewing the application of the exclusion criteria “active fault zones” according to the Site Selection Act, the project KaStör reviewed the current knowledge on active faults and fault zones in Germany and recommends methodological approaches to date and identify the activity of faulting. For the time being, the Site Selection Act defines 100 ∘C as a draft limit on the temperature at the outer surface of a repository container for all host rocks. The project Grenztemperatur studied the temperature dependency of the different thermal-hydraulic-mechanical-chemical/biological (THMC/B) processes according to available features-events-processes (FEP) catalogues for rock salt, clay stone, and crystalline rock and describes ways to defining host rock specific maximum temperatures based on specific disposal and safety concepts. Safety oriented weighting of different criteria and comparison of different potential regions and sites are key challenges during the siting process. The project MABeSt studied and reviewed methodological approaches to this weighting and comparison problem with special emphasis on multi criteria analysis (MCA) and multi criteria decision analysis (MCDA). A key requirement for safe geological disposal of nuclear waste is barrier integrity. The project PeTroS performed the first triaxial flow-through experiments on natural rock salt samples at disposal relevant p−T conditions and studied potential percolation mechanisms of fluids within rock salt. The data substantiate that the minimum stress criterion and/or the dilatancy criterion are the prime “percolation thresholds” in rock salt. The research results support BASE in fulfilling its tasks as national regulator according to state-of-the-art science and technology and are also relevant to other stakeholders of the siting process.

scholarly journals Leaching Savannah River Plant Nuclear Waste Glass in a Saturated Tuff Environment

1984 ◽  
Vol 44 ◽  
Author(s):  
N. E. Bibler ◽  
G. G. Wicks ◽  
V. M. Oversby
Keyword(s):  
Stainless Steel ◽  
Radioactive Waste ◽  
Nuclear Waste ◽  
304L Stainless Steel ◽  
Savannah River ◽  
Mass Losses ◽  
Savannah River Plant ◽  
Simulated Waste ◽  
Nuclear Waste Glass ◽  
Time Periods

AbstractSamples of SRP glass containing either simulated or actual radioactive waste were leached at 90°C under conditions simulating a saturated tuff repository environment. The leach vessels were fabricated of tuff and actual tuff groundwater was used. Thus, the glass was leached only in the presence of those materials (including the Type 304L stainless steel canister material) that would be in the actual repository. Tests were performed for time periods up to 6 months at a SA/V ratio of 100 m−1. Results with glass containing simulated waste indicated that stainless steel canister material around the glass did not significantly affect the leaching. Based on Li and B (elements not in significant concentrations in the tuff or tuff groundwater), glass containing simulated waste leached identically to glass containing actual radioactive waste. The tuff buffered the pH so that only a slight increase was observed as a result of leaching. Results with glass containing actual radioactive waste indicated that tuff reduced the concentrations of Cs-137, Sr-90, and Pu-238 in the free groundwater in the simulated repository by 10–100X. Also, radiolysis of the groundwater by the glass (approximately 1000 rad/hr) did not significantly affect the pH in the presence of tuff. Measured normalized mass losses in the presence of tuff for the glass based on Cs-137, Sr-90, and Pu-238 in the free groundwater were extremely low, nominally 0.02, 0.02, and 0.005 g/m2, respectively, indicating that the glass-tuff system retained radionuclides well.

Nature and Effect of the Alteration Layer During Nuclear Waste Glass Dissolution

2002 ◽  
Vol 713 ◽  
Author(s):  
A. Gauthier ◽  
P. Le Coustumer ◽  
J-H. Thomassin
Keyword(s):  
Nuclear Waste ◽  
Pure Water ◽  
Aqueous Media ◽  
Glass Dissolution ◽  
Icp Ms ◽  
Nuclear Waste Glass ◽  
Texture Structure ◽  
The Stability ◽  
Short Time

ABSTRACTThe goal of this study is to understand the role of the interface developed during R7T7 glass alteration. This glass has been leached in two different aqueous media (pure water, silica rich water and phosphorous rich water). The lixiviation tests have been optimized to assess the role of the alteration layer developed on the surface of the glass. The solution and the solid have been characterized by ICP-MS and TEM/X-EDS respectively. The results put in evidence that a complex alteration layer is formed. Its texture, structure and chemistry are discussed with respect to the evolution of the solution during the tests. The alteration layer is always present on the surface of the glass and is considered to control (at short time) diffusion of the different species through the layer. Further study must be undertaken to assess the evolution and the stability of the interface for longer time periods.

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