PNL-Sandia HLW Package Interactions Test: Phase One

1981 ◽  
Vol 6 ◽  
Author(s):  
Martin A. Molecke ◽  
Donald J. Bradley ◽  
John W. Shade
Keyword(s):  
Synergistic Effects ◽  
Fission Products ◽  
Significant Degree ◽  
Test System ◽  
High Level Waste ◽  
Uniform Corrosion ◽  
Surface Alteration ◽  
High Level ◽  
Bedded Rock Salt ◽  
Bedded Rock

ABSTRACTThe first phase of a complex high-level waste (HLW) package interactions test in a salt environment has been completed. The test system consisted of PNL 76–68 HLW glass (loaded with inactive fission products and 238U) surrounded by a stainless steel waste canister, a TiCode- 12 overpack, a bentonite/sand backfill, excess brine leachant, and a bedded rock salt container, all heldwithin a 19-liter autoclave. All components were physically compromised in order to force wasteform-barrier-salt interactions to occur during this 95-day, 250°C overtest. Analyses of leachant, wasteform, and all barrier surfaces were performed. Test data included the synergistic effects between barriers and confirmed previous analyses of simpler systems. The glass wasteform exhibited some surface alteration but was not dissolved to any significant degree. The TiCode-12 overpack showed minimal uniform corrosion and no localized attack. Observed mineralogical alteration of the backfill was minimal.

Vapour Phase Hydration of Blended Oxide - Magnox Waste Glasses

2003 ◽  
Vol 807 ◽  
Author(s):  
Neil C. Hyatt ◽  
William E. Lee ◽  
Russell J. Hand ◽  
Paul K. Abraitis ◽  
Charlie R. Scales
Keyword(s):  
Glass Matrix ◽  
Alkaline Earth ◽  
Vapour Phase ◽  
Waste Glass ◽  
High Level Waste ◽  
Zirconium Silicate ◽  
Constant Rate ◽  
Surface Alteration ◽  
High Level ◽  
Short Incubation Time

ABSTRACTVapour phase hydration studies of a blended Oxide / Magnox simulant high level waste glass were undertaken at 200°C, over a period of 5 – 25 days. The alteration of this simulant waste glass is characterised by a short incubation time of less than 5 days, leading to the formation of an alteration layer several microns thick. Following the incubation period, the alteration proceeds at a constant rate of 0.15(1)μmd−1. The distribution of key glass matrix (Si, Na) and waste (Cs, Zr, Nd, Mo) elements was found to vary significantly across the alteration layer. Vapour phase hydration leads to formation of surface alteration products, identified as smectite, zirconium silicate and alkaline-earth molybdate phases.

scholarly journals Characterization of Composite Ceramic High Level Waste Forms

1997 ◽  
Vol 481 ◽  
Author(s):  
S. M. Frank ◽  
K. J. Bateman ◽  
T. DiSanto ◽  
S. G. Johnson ◽  
T. L. Moschetti ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Composite Ceramic ◽  
Waste Form ◽  
High Level Waste ◽  
Ceramic Waste ◽  
Chloride Form ◽  
Physical Measurements ◽  
High Level

ABSTRACTArgonne National Laboratory has developed a composite ceramic waste form for the disposition of high level radioactive waste produced during electrometallurgical conditioning of spent nuclear fuel. The electrorefiner LiCl/KCl eutectic salt, containing fission products and transuranics in the chloride form, is contacted with a zeolite material which removes the fission products from the salt. After salt contact, the zeolite is mixed with a glass binder. The zeolite/glass mixture is then hot isostatic pressed (HIPed) to produce the composite ceramic waste form. The ceramic waste form provides a durable medium that is well suited to incorporate fission products and transuranics in the chloride form. Presented are preliminary results of the process qualification and characterization studies, which include chemical and physical measurements and product durability testing, of the ceramic waste form.

scholarly journals Development of a ceramic waste form for high-level waste disposal

1999 ◽  
Vol 556 ◽  
Author(s):  
D. W. Esh ◽  
K. M. Goff ◽  
K. T. Hirsche ◽  
T. J. Battisti ◽  
M. F. Simpson ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
High Level Waste ◽  
Durability Test ◽  
Ceramic Waste ◽  
High Level ◽  
The Impact

AbstractA ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

Characteristics and quantities of radioactive wastes

1986 ◽  
Vol 319 (1545) ◽  
pp. 5-16 ◽  
Keyword(s):  
Nuclear Power ◽  
Small Volume ◽  
Fuel Cycle ◽  
Residual Activity ◽  
Fission Products ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Cladding ◽  
Waste Streams ◽  
High Level

Estimates are given of the total quantities of radioactivity, and of the contribution from different isotopes to this total, arising in the wastes from civil nuclear power generation; the figures are normalized to 1 GW (e) y of power production. The intensity of the heat and y-radiation emitted by the spent fuel has been calculated, and their decrease as the radioactivity decays. Reprocessing the spent fuel results in 95% or more of the fission products and higher actinides being concentrated in a small volume of high-level, heat-emitting waste. The total decay curve of unreprocessed spent fuel or of the separated high-level waste is dominated by the decay of some fission products for a few hundred years and then by the decay of some actinide isotopes for some tens of thousands of years. The residual activity is compared with that of the original uranium ore. Some of the long-lived activity will appear in other waste streams, particularly on the fuel cladding, and the volumes and activities of these wastes arising in this country are recorded. The long-lived activity arising from reactor decommissioning will be small compared with the annual arisings from the fuel cycle.

Hot Cell Preparation of Tesiing Materials

1984 ◽  
Vol 44 ◽  
Author(s):  
R. E. Thornhill ◽  
C. A. Knox
Keyword(s):  
Fission Products ◽  
Materials Characterization ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Cell Preparation ◽  
High Radiation ◽  
High Level ◽  
Commercial Reactor ◽  
Simulated High Level Waste

AbstractIt is important in nuclear waste repository development that testing be done with materials containing a radionuclide spectrum representative of actual wastes. To meet the need for such materials, the Materials Characterization Center (MCC) has prepared simulated high level waste (HLW) glasses with radionuclides representative of about 10-, 300-, and 1000-year-old waste. A quantity of well characterized spent fuel also has been acquired for the same purpose. Glasses containing 10- and 300-year-old wastes, and the spent fuel specimens, must be fabricated in a hot cell. Hot cell conditions (high radiation field, remote operation, and difficulty of repairs) require that procedures and equipment normally used in materials preparation out-of-cell be modified for hot cell applications.This paper discusses the fabrication of two glasses, and the preparation of test specimens of these glasses and spent fuel. One of the glasses is a 76–68 composition, which is fully loaded with actual commercial reactor fission product waste. The other glass contains simulated Barnwell Nuclear Fuel Plant waste, doped with different combinations of fission products and actinides. The spent fuel is a 10-year-old PWR material. Special techniques have been used to achieve high quality, well characterized testing materials, including specimens in the form of segments, wafers, cylinders, and powders of these materials.

ThermoChimie, the ANDRA Thermodynamic Database

2012 ◽  
Vol 1475 ◽  
Author(s):  
Lara Duro ◽  
Mireia Grivé ◽  
Eric Giffaut
Keyword(s):  
Performance Assessment ◽  
Thermodynamic Data ◽  
Fission Products ◽  
Major Elements ◽  
High Level Waste ◽  
Experimental Program ◽  
Waste Repository ◽  
High Level ◽  
Inorganic Ligands ◽  
Geological Formations

ABSTRACTThermoChimie is the thermodynamic data base initially created by ANDRA in 1996, especially designed and qualified for systems of interest for the French high level waste repository concept. This database is supported by an experimental program on actinides and fission products and also on major components of the systems of interest, and it has been continuously updated since its creation. The validation of the database is continuously on-going through geochemical calculations related to the performance assessment of different backfill/buffer materials and/or geological formations. ThermoChimie contains data on major elements (including stability of minerals such as clays, zeolites, cementitious phases), a long list of radioelements, such as actinides and lanthanides, chemotoxic metals, as well as organic and inorganic ligands.

Research Priorities for Nuclear Waste Isolation

1981 ◽  
Vol 11 ◽  
Author(s):  
T.H. Pigford
Keyword(s):  
Synergistic Effects ◽  
Research Priorities ◽  
High Level Waste ◽  
Crystal Formation ◽  
List Type ◽  
Processing Technologies ◽  
Nondestructive Examination ◽  
Glass Melter ◽  
High Level ◽  
Technical Basis

Hr. Gordon's comments are specific to defense waste only. The research priorities center around:1.Improvements in the technology for immobilizing high-level waste. For waste vitrification, these improvements include lower corrosiveness, less crystal formation, lower volatility, reduction in glass cracking, increased life of glass-melter, nondestructive testing methods, and remote handling techniques.2.Better understanding of the performance of the waste package components in the repository environment. This requires additional research in the areas of repository characterization, the mechanisms of waste form corrosion, radionuclide migration in geologic media, predictive models for projecting waste system performance.. and synergistic effects of waste system components.3.Confirmation of waste disposal risk assessments and their establishment as the technical basis for regulatory criteria.4.Research priorities for transuranic (TRU) waste isolation are focused on the development of processing technologies for immobilization and of improved instrumentation for classification and nondestructive examination.

scholarly journals Natural Clay Minerals as a Starting Material for Matrices for the Immobilization of Radioactive Waste from Pyrochemical Processing of SNF

2021 ◽  
Vol 13 (19) ◽  
pp. 10780
Author(s):  
Anna V. Matveenko ◽  
Andrey P. Varlakov ◽  
Alexander A. Zherebtsov ◽  
Alexander V. Germanov ◽  
Ivan V. Mikheev ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Storage Time ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Ceramic Materials ◽  
High Level Waste ◽  
Natural Clay ◽  
High Level

Pyrochemistry is a promising technology that can provide benefits for the safe reprocessing of relatively fresh spent nuclear fuel with a short storage time (3–5 years). The radioactive waste emanating from this process is an electrolyte (LiCl–KCl) mixture with fission products included. Such wastes are rarely immobilized through common matrices such as cement and glass. In this study, samples of ceramic materials, based on natural bentonite clay, were studied as matrices for radioactive waste in the form of LiCl–KCl eutectic. The phase composition of the samples, and their mechanical, hydrolytic, and radiation resistance were characterized. The possibility of using bentonite clay as a material for immobilizing high-level waste arising from pyrochemical processing of spent nuclear fuel is further discussed in this paper.

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