scholarly journals Problematic Issues Regarding the Quantity and Properties of Vitrified High-Level Waste Returning to Ukraine

2019 ◽  
pp. 26-29
Author(s):  
Yu. Olkhovyk
Keyword(s):  
Glass Matrix ◽  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Slow Cooling ◽  
Structural Homogeneity ◽  
Aluminophosphate Glass ◽  
Long Term Storage ◽  
Technical Requirements ◽  
Order Increase ◽  
High Level

Safety justification of long-term storage and further disposal of vitrified high-level waste returning to Ukraine shall be based on reliable information about their physical and chemical characteristics, which include not only the radionuclide composition, but also the estimated evolution of Na-Al-P glass properties in the conditions of potential longterm effect of unfavorable factors. The paper indicates an inconsistency of dose coefficients, which according to the Energoatom standards shall be used to calculate the amount of high-level waste returning to Ukraine after storage and processing of VVER-440 spent nuclear fuel, with the regulatory requirements of the country supplying vitrified high-level waste. The quantitative assessment of transuranium radionuclides and technetium 99 entering the glass matrix also requires a critical review. The research considered the possibility of uncertainty related to the structural homogeneity of a glass matrix due to the underestimation of cracking and crystallization processes that occur in the package in sodium-aluminophosphate glass cooling. The presence of a large number of rare-earth oxides in sodium-aluminophosphate glass contributes to its crystallization in slow cooling with monazite formation. These phenomena can lead to a partial conversion of amorphous glass into a crystalline phase accompanied by 1-2 order increase in the velocity of leaching of elements. When developing technical requirements for vitrified high-level waste returning to Ukraine, it is necessary to insist on the provision of experimentally determined parameters of the structural homogeneity of glass blocks. There is a need for obtaining experimentally defined parameters of radiation resistance of a sodium-aluminophosphate matrix under the influence of a dose that can be accumulated over a period of 100 years using accelerated self-radiation methods.

Deep Geological Disposal of Spent Nuclear Fuel and High-Level Waste

2015 ◽  
pp. 367-399
Author(s):  
Želimir Veinović ◽  
Biljana Kovačević Zelić ◽  
Dubravko Domitrović
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Modern World ◽  
Geological Disposal ◽  
Long Term Storage ◽  
Deep Geological Disposal ◽  
Safety Cases ◽  
High Level

Management of Spent Nuclear Fuel (SF) and High-Level Waste (HLW) is one of the most important and challenging problems of the modern world. Otherwise a clean, cheap, constant, and secure way to produce electricity, nuclear power plants create large amounts of highly hazardous waste. Repositories—deep Geological Disposal Facilities (GDF)—for these types of waste must prevent radionuclides from reaching the biosphere, for up to 1,000,000 years, migrating from a deep (more than 300m), stable geological environment. At present, there are no operating GDFs for SF and/or HLW, mostly due to the difficult and complex task of preparing safety cases and licensing. The purpose of this chapter is to validate the generic R&D activities in this area and present alternative concepts of Radioactive Waste (RW) management: retrievability, reversibility, regional GDFs, long-term storage, and deep borehole disposal, demonstrating the main engineering tasks in solving the problem of RW management and disposal.

Copper Valence and Local Environment in Aluminophosphate Glass-Ceramics for Immobilization of High Level Waste from Uranium-Graphite Reactor Spent Nuclear Fuel Reprocessing

2015 ◽  
Vol 1744 ◽  
pp. 79-84 ◽  
Author(s):  
Sergey V. Stefanovsky ◽  
Andrey A. Shiryaev ◽  
Michael B. Remizov ◽  
Elena A. Belanova ◽  
Pavel A. Kozlov ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Glass Ceramics ◽  
Local Environment ◽  
Degree Of Crystallinity ◽  
High Level Waste ◽  
Spent Fuel ◽  
Aluminophosphate Glass ◽  
Copper Valence ◽  
High Level ◽  
Aluminophosphate Glasses

ABSTRACTCopper valence and environment in two sodium aluminophosphate glasses suggested for immobilization of HLW from reprocessing of spent fuel of uranium-graphite channel reactor (Russian AMB) were studied by XRD, SEM/EDX, XAFS and EPR. Target glass formulations contained ∼2.4-2.5 mol.% CuO. The quenched samples were predominantly amorphous. The annealed MgO free sample had higher degree of crystallinity than the annealed MgO-bearing sample but both them contained orthophosphate phases. Cu in the materials was partitioned in favor of the vitreous phase. In all the samples copper is present as major Cu2+ and minor Cu+ ions. Cu2+ ions form planar square complexes (CN=4) with a Cu2+-O distance of 1.93-1.95 Å. Two more ions are positioned at a distance of 2.76-2.86 Å from Cu2+ ions. So the Cu2+ environment looks like a strongly elongated octahedron as it also follows from the absence of the pre-edge peak due to 1s→3d transition in Cu K edge XANES spectra of the materials. Cu+ ions form two collinear bonds at Cu+-O distances of 1.80-1.85 Å. Thus average Cu coordination number (CN) in the first shell was found to be 2.7-3.0.

Safety and Sustainability Implications of Long Term Storage of Radioactive Waste

2003 ◽  
Author(s):  
John Rowat
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Information Transfer ◽  
Spent Nuclear Fuel ◽  
Position Paper ◽  
High Level Waste ◽  
Long Term Storage ◽  
High Level ◽  
Term Storage

Storage and disposal of radioactive waste are complementary rather than competing activities, and both are required for the safe management of wastes. Storage has been carried out safely within the past few decades, and there is a high degree of confidence that it can be continued safely for limited periods of time. However, as the amounts of radioactive waste in surface storage have increased, concern has grown over the sustainability of storage in the long term and the associated safety and security implications. In response to these concerns, the IAEA has prepared a position paper [1] that is intended for general readership. This presentation will provide a summary of the position paper, and a discussion of some safety issues for further consideration. A key theme is the contrast of the safety and sustainability implications of long term storage with those of early disposal. A number of factors are examined from different points of view, factors such as safety and security, need of maintenance, institutional control and information transfer, community attitudes and availability of funding. The timing and duration of the process of moving from storage to disposal, which are influenced by factors such as the long timeframes required to implement disposal and changing public attitudes, will also be discussed. The position paper focuses on the storage of three main types of waste: high level waste from the reprocessing of nuclear fuel, spent nuclear fuel that is regarded as waste and long-lived intermediate level radioactive waste. Long term storage of mining and milling waste, and other large volumes of waste from processes involving the use of naturally occurring radioactive materials are not discussed. Specialist meetings were held last year by the IAEA on the sustainability and safety of long-term storage to establish and discuss the issues where a broad consensus exists, and to investigate areas where issues remain unresolved. Within the technical community, it is widely agreed that perpetual storage is not considered to be either feasible or acceptable because of the impossibility of assuring active control over the time periods for which these wastes remain potentially hazardous. For high-level and long-lived radioactive waste, the consensus of the waste management experts is that disposal in deep underground engineered facilities — geological disposal — is the best option that is currently available, or likely to be available in the foreseeable future.

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.

Preparation and characterization of vitrified glass matrix for high level waste from MOX fuel processing

2007 ◽  
Vol 353 (52-54) ◽  
pp. 4647-4653 ◽  
Author(s):  
V.S. Yalmali ◽  
D.S. Deshingkar ◽  
P.K. Wattal ◽  
S.R. Bharadwaj
Keyword(s):  
Glass Matrix ◽  
High Level Waste ◽  
Fuel Processing ◽  
Mox Fuel ◽  
High Level

scholarly journals Reactive transport calculations to evaluate sulphide fluxes in the near-field of a SF/HLW repository

2019 ◽  
Vol 98 ◽  
pp. 10005
Author(s):  
Marek Pękala ◽  
Paul Wersin ◽  
Veerle Cloet ◽  
Nikitas Diomidis
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Waste Containment ◽  
Geological Repository ◽  
High Level ◽  
Reducing Bacteria ◽  
Hlw Repository

Radioactive waste is planned to be disposed in a deep geological repository in the Opalinus Clay (OPA) rock formation in Switzerland. Cu coating of the steel disposal canister is considered as potential a measure to ensure complete waste containment of spent nuclear fuel (SF) and vitrified high-level waste (HLW) or a period of 100,000 years. Sulphide is a potential corroding agent to Cu under reducing redox conditions. Background dissolved sulphide concentrations in pristine OPA are low, likely controlled by equilibrium with pyrite. At such concentrations, sulphide-assisted corrosion of Cu would be negligible. However, the possibility exists that sulphate reducing bacteria (SRB) might thrive at discrete locations of the repository’s near-field. The activity of SRB might then lead to significantly higher dissolved sulphide concentrations. The objective of this work is to employ reactive transport calculations to evaluate sulphide fluxes in the near-field of the SF/HLW repository in the OPA. Cu canister corrosion due to sulphide fluxes is also simplistically evaluated.

Interim Storage Technology of Spent Fuel and High-Level Waste in Germany

2007 ◽  
Author(s):  
H. Geiser ◽  
J. Schro¨der
Keyword(s):  
Normal Operation ◽  
High Level Waste ◽  
Spent Fuel ◽  
Horizontal Orientation ◽  
Long Term Storage ◽  
Fuel Storage ◽  
Safety Features ◽  
High Level ◽  
Interim Storage ◽  
Aircraft Crash

The idea of using casks for interim storage of spent fuel arose at GNS after a very controversial political discussion in 1978, when total passive safety features (including aircraft crash conditions) were required for an above ground spent fuel storage facility. In the meantime, GNS has loaded more than 1000 casks at 25 different storage sites in Germany. GNS cask technology is used in 13 countries. Spent fuel assemblies of PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high level waste containers are stored in full metal casks of the CASTOR® type. Also MOX fuel of PWR and BWR has been stored. More than two decades of storage have shown that the basic requirements (safe confinement, criticality safety, sufficient shielding and appropriate heat transfer) have been fulfilled in any case — during normal operation and in case of severe accidents, including aircraft crash. There is no indication of problems arising in the future. Of course, the experience of more than 20 years has resulted in improvements of the cask design. The CASTOR® casks have been thoroughly investigated by many experiments. There have been approx. 50 full and half scale drop tests and a significant number of fire tests, simulations of aircraft crash, investigations with anti tank weapons, and an explosion of a railway tank with liquid gas neighbouring a loaded CASTOR® cask. According to customer and site specific demands, different types of storage facilities are realized in Germany. Firstly, there are facilities for long-term storage, such as large ventilated central storage buildings away from reactor or ventilated storage buildings at the reactor site, ventilated underground tunnels or concrete platforms outside a building. Secondly, there are facilities for temporary storage, where casks have been positioned in horizontal orientation under a ventilated shielding cover outside a building.

Adsorption Materials Development for the Separation of Actinides and Specific Fission Products from High Level Waste

2014 ◽  
Vol 94 ◽  
pp. 103-110 ◽  
Author(s):  
Yue Zhou Wei ◽  
Shun Yan Ning ◽  
Qi Long Wang ◽  
Zi Chen ◽  
Yan Wu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Selective Adsorption ◽  
Packed Column ◽  
Liquid Waste ◽  
Separation Method ◽  
High Level Waste ◽  
Small Scale ◽  
Fast Diffusion ◽  
High Level Liquid Waste ◽  
High Level

The long-term radiotoxicity of high level liquid waste (HLLW) generated in spent nuclear fuel reprocessing is governed by the content of several long-lived minor actinides (MA) and some specific fission product nuclides. To efficiently separate MA (Am, Cm) and some FPs such as Cs and Sr from the HLLW, we have been studying an advanced aqueous partitioning process, which uses selective adsorption as separation method. In this work, we prepared different types of porous silica-based organic/inorganic adsorbents with fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. So they are advantageously applicable to efficient separation of the MA and specific FP elements from HLLW. Adsorption and separation behaviors of the MA and some FP elements such as Cs and Sr were studied. Small scale separation tests using simulated and genuine nuclear waste solutions were carried out and the obtained results indicate that the proposed separation method based on selective adsorption is essentially feasible.

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.

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