Technical Workshop on Near-Field Performance Assessment of High-Level Waste Disposal

1993 ◽  
Vol 333 ◽  
Author(s):  
Michael Apted ◽  
Kjell Andersson
Keyword(s):  
Nuclear Energy ◽  
Source Term ◽  
International Workshop ◽  
Near Field ◽  
Field Performance ◽  
High Level Waste ◽  
International Survey ◽  
Energy Agency ◽  
High Level ◽  
Waste Repositories

ABSTRACTAn international Workshop was conducted to discuss the present state of modelling of near-field performance of high-level waste repositories, with emphasis on the conceptual and mathematical models for source-term calculations. The Workshop was based on an international survey and review of this topic by the authors. Technical position papers developed by the participants are summarized here. A Proceedings of the Workshop, including the full position papers and the survey and review document, are to be published by Nuclear Energy Agency by the end of 1993.

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.

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.

Near Field Radionuclide Concentrations: Sorption or Solubility Constrained?

2001 ◽  
Author(s):  
Bernhard Kienzler ◽  
Peter Vejmelka ◽  
Volker Metz
Keyword(s):  
Portland Cement ◽  
Performance Assessment ◽  
Geochemical Modeling ◽  
Coal Fly Ash ◽  
Near Field ◽  
High Level Waste ◽  
Sorption Behavior ◽  
Geochemical Environment ◽  
Backfill Materials ◽  
High Level

Abstract The amount of mobile radionuclides is controlled by the geochemical isolation potential of the repository. Many investigations are available in order to determine the maximal radionuclide concentrations released from different waste forms of specific disposal strategies for disposal in rock salt formations. These investigations result in reaction (dissolution) rates, maximum concentrations, and sorption coefficients. The experimental data have to be applied to various disposal strategies. The case studies presented in this communication cover the selection, the volumes, and the composition of backfill materials used as sorbents for radionuclides. As an example, for brown coal fly ash (BFA) - Q-brine systems, sorption coefficients were measured as well as solublilities of several actinides and other long-lived radionuclides. Dissolved CO32− was buffered to negligible concentration by the presence of high amount of Mg in solution. In the sorption experiments Pu, Th, Np, and U concentrations close or below detection limit were obtained. Concentrations in the same ranges are computed by means of geochemical modeling, if precipitation of “simple” tetravalent hydroxides (An(OH)4(am) phases) is assumed. In the case of U in a Portland cement dominated geochemical environment, measured U(VI) concentration corresponds to the solubility of hexavalent solids, such as Na2U2O7. A similar behavior of U was observed in high-level waste glass experiments. Experiments investigating sorption behavior of corroded cement showed that in the case of application of a sufficient large inventory of actinides, measured concentrations were found to be independent of the inventory. In this case, measured concentrations were controlled by solid phases. If smaller actinide inventories were applied, resulting concentrations were found to be below concentrations constrained by well-known solids. Here, a more or less pronounced sorption of the radioactive elements was observed. The radionuclide concentrations determined in the BFA “sorption” experiments are found to be close to the detection limits. For this reason, it is not possible to extrapolate the radionuclide behavior to lower concentrations. We cannot distinguish, if sorption or precipitation controls measured radionuclide concentrations. However, in the presence of reducing materials such as BFA, solubilities of tetravalent actinides and of Tc(IV) represent a realistic estimation of the maximal element concentrations needed in performance assessment studies. The concentrations of these redox sensitive elements are controlled by precipitation of An(OH)4(am) phases for disposal concepts considered in German salt formations. Under this assumption, quantities such as solid-solution ratios used in (sorption) experiments do not affect the mobilization of the radionuclides. Additional conclusions can be drawn from comparison of the findings for the redox sensitive elements in the BFA / portland cement brine systems: We can assume that expected actinide and technetium concentrations in the near-field of radioactive wastes are affected by the total inventory of radionuclides in the disposal room. Sorption will be relevant, if the total dissolved radionuclide concentration remains below the maximal solubility defined by the solid radionuclide phase which is stable in the geochemical environment. In contrast to the portland cement system, the relevant radionuclide phase are most probably tetravalent hydroxides in the BFA systems. These conclusions are of high importance to performance assessment for the radioactive waste repository systems, because they restrict the applicability of sorption models in the near field of the waste.

Optimal mixing scheme for graphite–bentonite mixtures used as buffer materials in high-level waste repositories

2021 ◽  
Vol 80 (17) ◽  
Author(s):  
Yun-zhi Tan ◽  
Zi-yang Xie ◽  
Fan Peng ◽  
Fang-hong Qian ◽  
Hua-jun Ming
Keyword(s):  
High Level Waste ◽  
High Level ◽  
Waste Repositories ◽  
Optimal Mixing

Modelling Radionuclide Speciation and Solubility Limits in The Near-Field of a Deep Repository

1985 ◽  
Vol 50 ◽  
Author(s):  
Hans Wanner
Keyword(s):  
Near Field ◽  
Cast Steel ◽  
Significant Rise ◽  
High Level Waste ◽  
Glass Waste ◽  
Carbonate Concentration ◽  
Realistic Description ◽  
Carbonate Complexation ◽  
Main Components ◽  
High Level

AbstractIn the safety analysis recently reported for a potential Swiss high-level waste repository, radionuclide speciation and solubility limits are calculated for expected granitic groundwater conditions. With the objective of deriving a more realistic description of radionuclide release from the near-field, an investigation has been initiated to quantitatively specify the chemistry of the near-field. In the Swiss case, the main components of the near-field are the glass waste-matrix, a thick cast steel canister horizontally stored in a drift, and a backfill of highly compacted bentonite.Based on available experimental data, an ion-exchange model for sodium, potassium, magnesium, and calcium has been developed, in order to simulate the reaction of sodium bentonite backfill with groundwater. The model assumes equilibrium with calcite as long as sufficient carbonates remain in the bentonite, as well as quartz saturation. The application of this model to the reference groundwater used in ‘Project Gewaehr 85’ results in a significant rise in pH (by up to 3 units) as well as a marked increase in the carbonate concentration.Neptunium and plutonium speciation and solubility limits are calculated for the reference groundwater chemistry gradually altered to that of saturated bentonite water and back again by a water exchange cycle model. The solubility limits estimated in this way generally turn out to be higher for the bentonite water than for the reference groundwater, mainly due to carbonate complexation of the actinide components AnO2+ and AnO22+. Uncertainties are particularly large for neptunium solubility due to its strong Eh dependence in bentonite water.

An Approach to Analysis of High Level waste Container Performance in an Unsaturated Repository Environment

1993 ◽  
Vol 333 ◽  
Author(s):  
John C. Walton ◽  
Narasi Sridhar ◽  
Gustavo Cragnolino ◽  
Tony Torng ◽  
Prasad Nair
Keyword(s):  
Dominant Role ◽  
Near Field ◽  
Temperature Fields ◽  
Localized Corrosion ◽  
High Level Waste ◽  
Quantitative Methodology ◽  
Waste Container ◽  
Wide Range ◽  
Engineered Barrier ◽  
High Level

ABSTRACTOne of the requirements for the performance of waste packages prescribed in 10CFR 60.113 is that the high level waste must be “substantially completely” contained for a minimum period of 300 to 1000 years. During this period, the radiation and thermal conditions in the engineered barrier system and the near-field environment are dominated by fission product decay. In the present U.S design of the engineered barrier system, the outer container plays a dominant role in maintaining radionuclide containment. A quantitative methodology for analyzing the performance of the container is described in this paper. This methodology enables prediction of the evolution of the waste package environment in terms of temperature fields, stability of liquid water on the container surface, and concentration of aggressive ions such as chloride. The initiation and propagation of localized corrosion is determined by the corrosion potential of the container material and critical potentials for localized corrosion. The coiTOsion potential is estimated from the kinetics of the anodic and cathodic reactions including oxygen diffusion through scale layers formed on the container surface. The methodology described is applicable to a wide range of metals, alloys and environmental conditions.

scholarly journals Is it Possible to Demonstrate Compliance with the Regulations for High-Level-Waste Repositories?

1992 ◽  
Vol 294 ◽  
Author(s):  
Felton W. Bingham
Keyword(s):  
Numerical Modeling ◽  
Data Collection ◽  
Performance Assessments ◽  
High Level Waste ◽  
Spent Fuel ◽  
Distant Future ◽  
Site Investigations ◽  
High Level ◽  
Waste Repositories ◽  
Shed Light

ABSTRACTThe regulations that currently govern repositories for spent fuel and high-level waste require demonstrations that are sometimes described as impossible to make. To make them will require an understanding of the current and the future phenomena at repository sites; it will also require credible estimates of the probabilities that the phenomena will occur in the distant future. Experts in many fields—earth sciences, statistics, numerical modeling, and the law—have questioned whether any amount of data collection can allow modelers to meet these requirements with enough confidence to satisfy the regulators. In recent years some performance assessments have begun to shed light on this question because they use results of actual site investigations. Although these studies do not settle the question definitively, a review of a recent totalsystem assessment suggests that compliance may be possible to demonstrate. The review also suggests, however, that the demonstration can be only at the “reasonable” levels of assurance mentioned, but not defined, in the regulations.

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