Leaching/Migration of UO2-Fuel in Compacted Bentonite

1989 ◽  
Vol 176 ◽  
Author(s):  
Y. Albinsson ◽  
R. Forsyth ◽  
G. Skarnemark ◽  
M. Skålberg ◽  
B. Torstenfelt ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
High Mobility ◽  
Bentonite Clay ◽  
Compacted Bentonite ◽  
Fuel Pellets ◽  
Powdered Copper ◽  
Trivalent Actinides ◽  
And Migration ◽  
Nuclear Fuel Pellets

ABSTRACTThe release and migration of the fission products cesium, europium and technetium, the actinides plutonium, americium and curium, and the activation product cobalt from spent nuclear fuel pellets in highly compacted bentonite clay has been measured after contact times of 101 and 386 days. Experiments at longer contact times are in progress. In some cases small amounts (0.5–1%) of powdered copper or iron metal, or vivianite (Fe3(PO4)2) have been mixed with the bentonite clay.The results indicate as expected a high mobility of cesium. The actinides have a very low mobility. After 386 days, plutonium has diffused less than 0.5 mm away from the fuel, while americium and curium appear to be somewhat more mobile. The behaviour of europium is similar to that of trivalent actinides. Very little technetium has been leached from all samples. Cobalt shows a strong retention in pure bentonite as well as in the presence of vivianite, while the mobility is much larger when iron or copper is added.

Ceramic Immobilization Options for Technetium

MRS Advances ◽  
2017 ◽  
Vol 2 (13) ◽  
pp. 753-758 ◽  
Author(s):  
Martin C. Stennett ◽  
Tae-Hyuk Lee ◽  
Daniel J. Bailey ◽  
Erik V. Johnstone ◽  
Jong Heo ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Thermo Gravimetric Analysis ◽  
Fission Products ◽  
High Mobility ◽  
Biologically Active ◽  
Purex Process ◽  
Gravimetric Analysis ◽  
X Ray Diffraction ◽  
Exchange Processes ◽  
High Concentrations

ABSTRACTLong half-life biologically active fission products, such as technetium-99, present particular problems for the disposal of spent nuclear fuel (SNF). Technetium is present in relatively high concentrations in fuel (approx. 1kg tonne-1 SNF) and has very high mobility in oxidizing environments. Technetium is therefore generally removed from SNF either by solvent extraction and reduction, during the PUREX process, or by sorption via ion exchange processes. Historically technetium has been disposed of via dilution and dispersion in the sea but stringent regulations now mean that the preferred long term option is immobilization in a highly stable and durable matrix. In this contribution we have looked at the synthesis of fluorite derivative crystalline host phases based on the zirconolite structure. Samples have been characterised by X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDX), thermo-gravimetric analysis (TG), and mass spectroscopy (MS). We have used Mo as an inactive surrogate for Tc.

Leaching of Spent Fuel under Anaerobic and Reducing Conditions

2002 ◽  
Vol 757 ◽  
Author(s):  
Yngve Albinsson ◽  
Arvid Ödegaard-Jensen ◽  
Virginia M. Oversby ◽  
Lars O. Werme
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Fluid Phase ◽  
Inert Atmosphere ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Reducing Conditions ◽  
Reaction Vessels

ABSTRACTSweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

scholarly journals Effects of Soluble Fission Products on Thermal Conductivities of Nuclear Fuel Pellets

1994 ◽  
Vol 31 (8) ◽  
pp. 796-802 ◽  
Author(s):  
Shinji ISHIMOTO ◽  
Mutsumi HIRAI ◽  
Kenichi ITO ◽  
Yoshiaki KOREI
Keyword(s):  
Nuclear Fuel ◽  
Fission Products ◽  
Fuel Pellets ◽  
Nuclear Fuel Pellets ◽  
Thermal Conductivities

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.

scholarly journals Simulation of alpha dose for predicting radiolytic species at the surface of spent nuclear fuel pellets

2014 ◽  
Vol 13 (1) ◽  
Author(s):  
Frank Becker ◽  
Bernhard Kienzler
Keyword(s):  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Strong Impact ◽  
Water Contact ◽  
Fuel Pellets ◽  
Particle Dose ◽  
Underground Disposal ◽  
Nuclear Fuel Pellets ◽  
Radionuclide Release

AbstractIn many countries, spent nuclear fuel is considered as a waste form to be disposed of in underground disposal. Under deep host rock conditions, a reducing environment prevails. In the case of water contact, long-term radionuclide release from the fuel depends on dissolution processes of the UO2 matrix. The dissolution rate of irradiated UO2 is controlled by oxidizing processes facilitated by dissolved species formed by alpharadiolysis of water in contact with spent nuclear fuel. To understand the effect of the radiation, the information of the dose rate at the surface of the fuel and its proximity is needed. α particles contribute strongly due to their high linear energy transfer. However, their dose rate and the energy deposition at the fuel surface are difficult to measure. Cylindrical fuel pellets as used in fuel rods show specific features, such as the rim zone, where a higher Pu concentration and a different porosity of the fuel matrix is present. The a particle dose rate was determined by simulations with the code MCNPX with focus on the rim zone of a pellet. As a result a 40% increased dose level in the rim zone exists in comparison to the center of a pellet. The potential dominant and inhomogeneous α-dose distribution is supposed to have a strong impact on radiolysis phenomena and in turn on an inhomogeneous dissolution of elements over the surface.

The SR 97 Safety Assessment of a KBS 3 Repository for Spent Nuclear Fuel – Overview, Review Comments and New Developments

2002 ◽  
Vol 713 ◽  
Author(s):  
Allan Hedin ◽  
Ulrik Kautsky ◽  
Lena Morén ◽  
Patrik Sellin ◽  
Jan-Olof Selroos
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Bentonite Clay ◽  
Research Programme ◽  
High Strength ◽  
Functional Requirements ◽  
Future Evolution ◽  
Site Investigations ◽  
Potential Safety

ABSTRACTIn preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the longterm safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock.The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion.The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a programme for site investigations, for formulating functional requirements on the repository's barriers, and for prioritisation of research.SR 97 has been reviewed both by an international group of OECD/NEA experts and by Swedish authorities. The NEA reviewers concluded that “SR 97 provides a sensible illustration of the potential safety of the KBS-3 concept”, and no issues were identified that need to be resolved prior to proceeding to the investigation of potential sites. The authorities' conclusions were in principal consistent with those of the NEA.Uncertainties and lack of knowledge in different areas identified in SR 97 have strongly influenced the contents and structure of SKBs most recent research programme, RD&DProgramme 2001.Since SR 97, the methodology for probabilistic consequence analyses have been further developed. Analytic approximations to the numerical transport models used in SR 97 have been developed. The new models have been used to extend the probabilistic calculations in SR 97.

scholarly journals Paradoxical role of high mobility group box 1 in glioma: a suppressor or a promoter?

2017 ◽  
Author(s):  
Richard A. Seidu ◽  
Min Wu ◽  
Zhaoliang Su ◽  
Huaxi Xu
Keyword(s):  
Molecular Mechanisms ◽  
High Mobility ◽  
Treatment Resistance ◽  
High Mobility Group ◽  
Tissue Invasion ◽  
Self Sufficiency ◽  
Glycation End Products ◽  
And Migration ◽  
Proliferation And Migration

Gliomas represent 60% of primary intracranial brain tumors and 80% of all malignant types, with highest morbidity and mortality worldwide. Although glioma has been extensively studied, the molecular mechanisms underlying its pathology remain poorly understood. Clarification of the molecular mechanisms involved in their development and/or treatment resistance is highly required. High mobility group box 1 protein (HMGB1) is a nuclear protein that can also act as an extracellular trigger of inflammation, proliferation and migration, through receptor for advanced glycation end products and toll like receptors in a number of cancers including gliomas. It is known that excessive release of HMGB1 in cancer leads to unlimited replicative potential, ability to develop blood vessels (angiogenesis), evasion of programmed cell death (apoptosis), self-sufficiency in growth signals, insensitivity to inhibitors of growth, inflammation, tissue invasion and metastasis. In this review we explore the mechanisms by which HMGB1 regulates apoptosis and autophagy in glioma. We also looked at how HMGB1 mediates glioma regression and promotes angiogenesis as well as possible signaling pathways with an attempt to provide potential therapeutic targets for the treatment of glioma.

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