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.

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.

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.

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.

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.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

2003 ◽  
Vol 807 ◽  
Author(s):  
Paul Wersin ◽  
Lawrence H. Johnson ◽  
Bernhard Schwyn
Keyword(s):  
Solid Phase ◽  
Near Field ◽  
Intermediate Level ◽  
Redox Conditions ◽  
High Level Waste ◽  
Redox Potentials ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
High Level ◽  
Waste Repositories

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

Sign in / Sign up

Export Citation Format

Share Document