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.

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.

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.

scholarly journals Thermal Conductivity of Korean Compacted Bentonite Buffer Materials for a Nuclear Waste Repository

Energies ◽  
2018 ◽  
Vol 11 (9) ◽  
pp. 2269 ◽  
Author(s):  
Seok Yoon ◽  
WanHyoung Cho ◽  
Changsoo Lee ◽  
Geon-Young Kim
Keyword(s):  
Thermal Conductivity ◽  
Near Field ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Decay Heat ◽  
Bentonite Buffer ◽  
Buffer Material ◽  
Overall Evaluation ◽  
High Level

Engineered barrier system (EBS) has been proposed for the disposal of high-level waste (HLW). An EBS is composed of a disposal canister with spent fuel, a buffer material, backfill material, and a near field rock mass. The buffer material is especially essential to guarantee the safe disposal of HLW, and plays the very important role of protecting the waste and canister against any external mechanical impact. The buffer material should also possess high thermal conductivity, to release as much decay heat as possible from the spent fuel. Its thermal conductivity is a crucial property since it determines the temperature retained from the decay heat of the spent fuel. Many studies have investigated the thermal conductivity of bentonite buffer materials and many types of soils. However, there has been little research or overall evaluation of the thermal conductivity of Korean Ca-type bentonite buffer materials. This paper investigated and analyzed the thermal conductivity of Korean Ca-type bentonite buffer materials produced in Gyeongju, and compared the results with various characteristics of Na-type bentonites, such as MX80 and Kunigel. Additionally, this paper suggests various predictive models to predict the thermal conductivity of Korean bentonite buffer materials considering various influential independent variables, and compared these with results for MX80 and Kunigel.

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.

Nepheline crystallization behavior in simulated high-level waste glasses

2019 ◽  
Vol 505 ◽  
pp. 215-224 ◽  
Author(s):  
Devon L. McClane ◽  
Jake W. Amoroso ◽  
Kevin M. Fox ◽  
Albert A. Kruger
Keyword(s):  
Crystallization Behavior ◽  
High Level Waste ◽  
High Level ◽  
Simulated High Level Waste

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.

The Effect of Waste Package Components on Radionuclides Released from Spent Fuel Under Hydrothermal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
Shirley A. Rawson ◽  
William L. Neal ◽  
James R. Burnell
Keyword(s):  
Low Carbon Steel ◽  
Fission Products ◽  
Nuclear Waste Repository ◽  
Low Carbon ◽  
Spent Fuel ◽  
Hydrothermal Conditions ◽  
Waste Package ◽  
Dissolved Species ◽  
Radionuclide Release ◽  
Time Invariant

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.

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