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.

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.

Packaging of High Burnup UO2 and MOX Spent Fuel Considering Temperature Restrictions in the Near Field of a Repository

1997 ◽  
Vol 506 ◽  
Author(s):  
D.F. McGinnes ◽  
J. W. Schneider
Keyword(s):  
Near Field ◽  
Management Program ◽  
High Level Waste ◽  
Heat Output ◽  
Spent Fuel ◽  
Ambient Temperatures ◽  
Heat Loading ◽  
High Burnup ◽  
Maximum Temperatures ◽  
High Level

ABSTRACTThe direct disposal of spent fuel is one of the options considered in the Swiss high level waste management program. One of the important questions, within this program, is the heat generation from high-burnup UO2and MOX spent fuels. Depending on the repository boundary conditions (e.g. ambient temperatures at depth, thermal properties of the host rock etc.), on the maximum temperatures allowed in the near field and on the heat output of the fuel, it may not always be possible to completely fill the conceptual waste canister. The aim of this paper is to address the potential loading of spent fuel into canisters for different possible repository heat loading restrictions

Laboratory and Modelling Studies of Sodium Benionite Grounmater Interaction

1986 ◽  
Vol 84 ◽  
Author(s):  
M. Sneujman ◽  
H. Uotiia ◽  
J. Rantanen
Keyword(s):  
Near Field ◽  
Chemical Species ◽  
Fluid Phase ◽  
High Level Waste ◽  
Sodium Bentonite ◽  
Model Calculations ◽  
Buffer Material ◽  
Modelling Studies ◽  
Calcite Saturation ◽  
High Level

AbstractAccording to the present Finnish concept sodium bentonite will be used as a buffer material in the repository for high-level waste. Experimental and theoretical studies treating the effect of bentonite upon the chemical conditions in a repository have been initiated with the object of specifying the chemistry of the near field.Sodium bentonite was let react with water under anaerobic conditions at 25°C for 540 days, during which time six fluid samples were extracted for the chemical analysis of 15 chemical species. The generated fluid phase was alkaline (PH = 9…10) and contained a high amount of bicarbonate. Also a low redox-potential was measured. The fluid phase chemistry was investigated using the geochemical code PHREEM. Calcite saturation was observed in all fluid samples.A modelling of sodium bentonite interaction with water based on the main mineral components of bentonite was also performed with PHREEQE. A fairly good agreement between experimental results and model calculations was observed.

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.

The Salton Sea Geothermal Field as a Natural Analog for the Near-Field in a Salt High-Level Nuclear Waste Repository

1984 ◽  
Vol 44 ◽  
Author(s):  
Wilfred A. Elders ◽  
Judith B. Moody
Keyword(s):  
Nuclear Waste ◽  
Near Field ◽  
Geothermal Field ◽  
Salton Sea ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Natural Analog ◽  
Naturally Occurring ◽  
Wide Range ◽  
High Level

AbstractThe Salton Sea Geothermal Field (SSGF), on the delta of the Colorado River in southern California, is being studied as a natural analog for the near-field environment of proposed nuclear waste repositories in salt. A combination of mineralogical and geochemical methods is being employed to develop a three dimensional picture of temperature, salinity, lithology, mineralogy, and chemistry of reactions between the reservoir rocks and the hot brines. Our aim is to obtain quantitative data on mineral stabilities and on mobilities of the naturally occurring radionuclides of concern in Commercial High-Level Waste (CHLW). These data will be used to validate the EQ3/6 geochemical code under development to model the salt near-field repository behavior.Maximum temperatures encountered in wells in the SSGF equal or exceed peak temperatures expected in a salt repository. Brines produced from these wells have major element chemistry similar to brines from candidate salt sites. Relative to the rocks, these brines are enriched in Na, Mn, Zn, Sr, Ra and Po, depleted in Ba, Si, Mg, Ti, and Al, and strongly depleted in U and Th. However the unaltered rocks contain only about 2–3 ppm of U and 4–12 ppm of Th, largely in detrital epidotes and zircons. Samples of hydrothermally altered rocks from a wide range of temperature and salinity show rather similar uniform low concentrations of these elements, even when authigenic illite, chlorite, epidote and feldspar are present. These observations suggest that U and Th are relatively immobile in these hot brines. However Ra, Po, Cs and Sr are relatively mobile. Work is continuing to document naturally occurring radionuclide partitioning between SSGF minerals and brine over a range of temperature, salinity, and lithology.

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.

Modelling sorption data for the actinides Am(III), Np(V) and Pa(V) on montmorillonite

2006 ◽  
Vol 94 (9-11) ◽  
Author(s):  
Michael H. Bradbury ◽  
B. Baeyens
Keyword(s):  
Near Field ◽  
High Level Waste ◽  
Spent Fuel ◽  
Data Set ◽  
Sorption Data ◽  
Deep Geological Disposal ◽  
Retention Characteristics ◽  
High Level ◽  
Trace Concentrations ◽  
First Time

The retention characteristics of the bentonite near-field engineered barrier proposed in most of the concepts for the deep geological disposal of high-level waste and spent fuel are an important component in repository performance assessment studies. Montmorillonite generally constitutes 65 to 90 wt.% of the bentonite. Sorption edge measurements have been performed at trace concentrations for the actinides Am(III), Np(V) and Pa(V) on purified and conditioned SWy-1 montmorillonite under anoxic, carbonate free conditions. To the best of the author´s knowledge, this is the first time a sorption data set has been measured for

The IAEA Coordinated Research Programme on the Performance of High-Level Waste Forms and Packages Under Repository Conditions

1992 ◽  
Vol 294 ◽  
Author(s):  
Vladimir S. Tsyplenkov
Keyword(s):  
Near Field ◽  
Research Programme ◽  
High Level Waste ◽  
Future Research ◽  
Spent Fuel ◽  
Waste Forms ◽  
Safety Assessments ◽  
High Level ◽  
Coordination Meeting

ABSTRACTThe IAEA initiated, in 1991, a Coordinated Research Programme (CRP), with the aim of promoting the exchange of information on the results obtained by different countries in the performance of high-level waste forms and waste packages under conditions relevant to final repository. These studies are being undertaken to obtain reliable data as input to safety assessments and environmental impact analyses, for final disposal purposes. The CRP includes studies on waste forms that are presently of interest worldwide: borosilicate glass, Synroc and spent fuel.Ten laboratories leading in investigation of high-level waste form performance have already joined the programme. The results of their studies and plans for future research were presented at the first Research Coordination Meeting, held in Karlsruhe, Germany, in November 1991. The technical contributions concentrated on effecting an understanding of dissolution mechanisms of waste forms under simulated repository conditions. A quantitative interpretation of the chemical processes in the near field is considered a prerequisite for long-term predictions and for the formulation of a "source term" for performance assessment studies.

scholarly journals Evaluation of temperature-induced effects on safety-relevant properties of clay host rocks with regard to HLW/SF disposal

2015 ◽  
Vol 79 (6) ◽  
pp. 1389-1395 ◽  
Author(s):  
M. Jobmann ◽  
A. Meleshyn
Keyword(s):  
Current Knowledge ◽  
Near Field ◽  
High Level Waste ◽  
Major Influence ◽  
Thermal Impact ◽  
Spent Fuel ◽  
Thermally Induced ◽  
Improved Performance ◽  
High Level ◽  
Host Rocks

AbstractDBE TECHNOLOGY, BGR and GRS are developing a methodology to demonstrate the safety of a repository for high-level waste and spent fuel (HLW/SF) in clays according to the requirements of the German regulating body. In particular, these requirements prescribe that the barrier effect of host rocks must not be compromised by a thermal impact resulting from HLW/SF emplacement. To substantiate and quantify this requirement, we carried out a literature survey of research on thermally-induced changes on clay properties. Effects thus compiled can be divided into thermo-hydro-mechanical and chemical-biological-mineralogical effects and were analysed with regard to their relevance to the integrity of clay host rocks. This analysis identified one effect of major influence within each group: thermal expansion and compaction as well as results of microbial activities. Importantly, it further revealed that a moderate temperature increase above 100°C cannot be expected to compromise the integrity of the geological barrier according to the current knowledge state. Evidence is presented in this paper that temperature increases up to 150°C can actually contribute to an improved performance of a radioactive waste repository by increasing the consolidation of the clay and sterilizing the repository's near-field to depress the deteriorative microbial effects. A quantitative temperature criterion for thermal impact of HLW/SF on clay host rocks is accordingly proposed.

Sign in / Sign up

Export Citation Format

Share Document