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.

Characteristics of Cementitious Paste for use in Deep Borehole Disposal of Spent Fuel and High Level Wasteforms

2015 ◽  
Vol 1744 ◽  
pp. 205-210 ◽  
Author(s):  
Nick C Collier ◽  
Karl P Travis ◽  
Fergus G F Gibb ◽  
Neil B Milestone
Keyword(s):  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Hardened Cement ◽  
Oil Well ◽  
Deep Borehole ◽  
Sodium Gluconate ◽  
Spent Fuel ◽  
High Temperature And Pressure ◽  
Temperature And Pressure ◽  
High Level

ABSTRACTDeep borehole disposal (or DBD) is now seen as a viable alternative to the (comparatively shallow) geologically repository concept for disposal of high level waste and spent nuclear fuel. Based on existing oil and geothermal well technologies, we report details of investigations into cementitious grouts as sealing/support matrices (SSMs) for waste disposal scenarios in the DBD process where temperatures at the waste package surface do not exceed ∼190ºC. Grouts based on Class G oil well cements, partially replaced with silica flour, are being developed, and the use of retarding admixtures is being investigated experimentally. Sodium gluconate appears to provide sufficient retardation and setting characteristics to be considered for this application and also provides an increase in grout fluidity. The quantity of sodium gluconate required in the grout to ensure fluidity for 4 hours at 90, 120 and 140°C is 0.05, 0.25 and 0.25 % by weight of cement respectively. A phosphonate admixture only appears to provide desirable retardation properties at 90°C. The presence of either retarder does not affect the composition of the hardened cement paste over 14 days curing and the phases formed are durable under conditions of high temperature and pressure.

Integrated Model of Korean Spent Fuel and High Level Waste Disposal Options

2009 ◽  
Author(s):  
Yongsoo Hwang ◽  
Ian Miller
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Near Field ◽  
Integrated Model ◽  
High Level Waste ◽  
Spent Fuel ◽  
Parameter Uncertainties ◽  
Design Concepts ◽  
High Level

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21’st century. The model addresses alternative design concepts for disposal of SNF of different types (CANDU, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model’s results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses.

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

2015 ◽  
Vol 1744 ◽  
pp. 73-78 ◽  
Author(s):  
Sergey V. Stefanovsky ◽  
Andrey A Shiryaev ◽  
Michael B. Remizov ◽  
Elena A. Belanova ◽  
Pavel A. Kozlov ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Local Environment ◽  
High Level Waste ◽  
Vitreous Phase ◽  
Xafs Spectroscopy ◽  
Graphite Reactor ◽  
State And Local ◽  
High Level ◽  
A Minor

ABSTRACTTwo Mo-bearing glasses considered as candidate forms for high level waste (HLW) a uranium-graphite reactor spent nuclear fuel (SNF) reprocessing were characterized. Incorporation of Mo in sodium aluminophosphate (SAP) glass increases its tendency to devitrification with segregation of orthophosphate phases. Valence state and local environment of Mo in the materials containing ∼2 wt.% MoO3 were determined by X-ray absorption fine structure (XAFS) spectroscopy. In the quenched samples composed of major vitreous and minor AlPO4 nearly all Mo is located in the vitreous phase as [Mo6+О6] units whereas in the annealed samples Mo is partitioned among vitreous and one or two orthophosphate crystalline phases in favor of the vitreous phase. Mo predominantly exists in a hexavalent state in distorted octahedral environment. Four oxygen ions are positioned at a distance of ∼1.71-1.73 Å and two - at a distance of 2.02-2.04 Å. Minor Mo(V) is also present as indicated by a response in EPR spectra with g ≈ 1.911-1.915.

scholarly journals German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
A. Schwenk-Ferrero
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
High Level Waste ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Cycle Simulation ◽  
High Level

Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content) and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104to 106years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

Salt waste treatment from a LiCl-KCl based pyrochemical spent fuel treatment process

2008 ◽  
Vol 96 (4-5) ◽  
Author(s):  
Mike T. Harrison ◽  
Howard E. Simms ◽  
Angela Jackson ◽  
Robert G. Lewin
Keyword(s):  
Alkali Metals ◽  
Treatment Process ◽  
Spent Nuclear Fuel ◽  
Waste Treatment ◽  
Electrochemical Techniques ◽  
Fission Products ◽  
High Level Waste ◽  
Spent Fuel ◽  
High Level ◽  
Salt Waste

Spent nuclear fuel may be treated using molten salt electrochemical techniques to separate fission products and actinide metals. Salt waste arising from the electrorefining process contains alkali metals, alkaline-earth and rare earth fission products, along with residual actinides. The removal of fission product elements has been investigated using zeolite ion exchange and phosphate precipitation, which allow the salt electrolyte to be recycled back into the main electrorefining vessel. Recycling the salt minimizes the volume of high level waste (HLW) generated and yields the fission products in a form more amenable to immobilization in a final disposal matrix. Several sets of experiments have been completed, all of which have significant implications for the use of these techniques on an industrial scale, as well as their ability to clean up the salt, and potentially produce robust and durable waste forms.

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.

TRU Recycling Options for Environmentally Friendly and Proliferation-Resistant Nuclear Fuel Cycle of FBR

2010 ◽  
Author(s):  
Sidik Permana ◽  
Mitsutoshi Suzuki
Keyword(s):  
High Performance ◽  
Fuel Cycle ◽  
Operation Time ◽  
Nuclear Nonproliferation ◽  
High Level Waste ◽  
Minor Actinide ◽  
Spent Fuel ◽  
Closed Fuel Cycle ◽  
The Core ◽  
High Level

The embodied challenges for introducing closed fuel cycle are utilizing advanced fuel reprocessing and fabrication facilities as well as nuclear nonproliferation aspect. Optimization target of advanced reactor design should be maintained properly to obtain high performance of safety, fuel breeding and reducing some long-lived and high level radioactivity of spent fuel by closed fuel cycle options. In this paper, the contribution of loading trans-uranium to the core performance, fuel production, and reduction of minor actinide in high level waste (HLW) have been investigated during reactor operation of large fast breeder reactor (FBR). Excess reactivity can be reduced by loading some minor actinide in the core which affect to the increase of fuel breeding capability, however, some small reduction values of breeding capability are obtained when minor actinides are loaded in the blanket regions. As a total composition, MA compositions are reduced by increasing operation time. Relatively smaller reduction value was obtained at end of operation by blanket regions (9%) than core regions (15%). In addition, adopting closed cycle of MA obtains better intrinsic aspect of nuclear nonproliferation based on the increase of even mass plutonium in the isotopic plutonium composition.

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

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