Functional Requirements and Performance Assessment: Application to the Case of Spent Fuel Disposal in Clay

2001 ◽  
Author(s):  
Jan Marivoet ◽  
Xavier Sillen ◽  
Peter De Preter
Keyword(s):  
Molecular Diffusion ◽  
Slow Release ◽  
Near Field ◽  
Functional Requirements ◽  
Spent Fuel ◽  
Systematic Analysis ◽  
Boom Clay ◽  
Safety Function ◽  
Clay Formation

Abstract Geological repository systems for the disposal of radioactive waste are based on a multi-barrier design. Individual barriers contribute in different ways to the overall long-term performance of the repository system, and furthermore, the contribution of each barrier can considerably change with time. In a systematic analysis of the functional requirements for achieving long-term safety a number of basic safety functions can be defined: physical confinement, retardation / slow release, dispersion / dilution and limited accessibility. In the case of the geological disposal of spent fuel in a clay formation a series of barriers are designed or chosen to contribute to the realisation of the basic safety functions. The physical confinement is realised by the watertight, high-integrity container, which prevents contact between groundwater and the confined radionuclides. In first instance the retardation / slow release function is realised by the slow dissolution of the waste matrix and by the limited solubility of many elements in the near field. However, the natural clay barrier provides the main contribution to this safety function. The migration of radionuclides through the Boom Clay is mainly due to molecular diffusion, which is an extremely slow process. Furthermore, many elements are strongly sorbed by the clay minerals what makes their migration even much slower. The dispersion / dilution function mainly occurs in the aquifer and the rivers draining the aquifer in the surroundings of the disposal system. Various performance indicators are used to quantify the contributions of each safety function and to explain the functioning of the repository system.

Static dissolution tests of α-doped UO2 under repository relevant conditions: Influence of Boom Clay and α-activity on fuel dissolution rates

2006 ◽  
Vol 932 ◽  
Author(s):  
Sonia Salah ◽  
Christelle Cachoir ◽  
Karel Lemmens ◽  
Norbert Maes
Keyword(s):  
Clay Fraction ◽  
Dissolution Kinetics ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Retention Capacity ◽  
Boom Clay ◽  
Dissolution Tests ◽  
Clay Formation ◽  
Α Activity

ABSTRACTSince reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ~263 µg · m−2 · d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.

Testing Safety and Performance Indicators in an Assessment of the Long-Term Performance of the Geological Disposal of Spent Fuel in Boom Clay

2003 ◽  
Author(s):  
Xavier Sillen ◽  
Jan Marivoet ◽  
Wim Cool ◽  
Peter de Preter
Keyword(s):  
Dose Rate ◽  
Performance Indicators ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Boom Clay ◽  
And Performance ◽  
The Individual ◽  
High Level ◽  
Clay Formation

The classical numerical output, or indicator, from assessments of the long-term safety of geological disposal systems for high-level radioactive waste is the individual effective dose rate. This indicator is an estimate of the possible individual health detriment and it is commonly compared to regulatory limits for assessing the safety of other nuclear activities as well, such as medical and industrial activities. As a safety indicator, the individual dose rate provides an estimate of the overall safety of the disposal system. However, because of the time frames involved in safety assessments of geological disposal systems, the need arises of complementary safety indicators that could be less affected by uncertainties like those associated with future human behaviour or the effects of climate change on the biosphere and the aquifers. Such alternative safety indicators can be, for example, radionuclide concentrations in the groundwater or fluxes to the biosphere due to a repository. Safety indicators only tell how globally safe a disposal system is. For confidence building, performance indicators can be used in addition to tell how the system works. In particular, performance indicators such as fluxes, activities or activity concentrations of selected radionuclides can show how the different components of the system fulfil their safety functions and contribute to the overall safety. The SPIN project of the European Commission assessed the usefulness of seven safety indicators and fourteen performance indicators by testing them in four actual assessments of disposal systems in granite formations. In this paper, indicators calculated from an assessment of the disposal of spent fuel in the poorly indurated Boom Clay formation are presented. Conclusions from the SPIN project that hold for repositories in clays are highlighted, as well as results that illustrate differences between the granite and clay disposal options. Finally, various performance and safety indicators are combined into a logical sequence to comprehensively present, and explain, the results of a safety assessment.

Static dissolution tests of α-doped UO2 under repository relevant conditions: Influence of Boom Clay and α-activity on fuel dissolution rates

2006 ◽  
Vol 932 ◽  
Author(s):  
Sonia Salah ◽  
Christelle Cachoir ◽  
Karel Lemmens ◽  
Norbert Maes
Keyword(s):  
Clay Fraction ◽  
Dissolution Kinetics ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Retention Capacity ◽  
Boom Clay ◽  
Dissolution Tests ◽  
Clay Formation ◽  
Α Activity

ABSTRACTSince reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ∼263 µg · m−2· d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.

The Spread of Pollutants from Harbour Sludge Depots

1984 ◽  
Vol 16 (3-4) ◽  
pp. 623-633
Author(s):  
M Loxham ◽  
F Weststrate
Keyword(s):  
Time Scales ◽  
Molecular Diffusion ◽  
Near Field ◽  
Parameter Analysis ◽  
Migration Patterns ◽  
Long Time ◽  
Dilution Effects ◽  
Short Time ◽  
Harbour Sludge

It is generally agreed that both the landfill option, or the civil techniques option for the final disposal of contaminated harbour sludge involves the isolation of the sludge from the environment. For short time scales, engineered barriers such as a bentonite screen, plastic sheets, pumping strategies etc. can be used. However for long time scales the effectiveness of such measures cannot be counted upon. It is thus necessary to be able to predict the long term environmenttal spread of contaminants from a mature landfill. A model is presented that considers diffusion and adsorption in the landfill site and convection and adsorption in the underlaying aquifer. From a parameter analysis starting form practical values it is shown that the adsorption behaviour and the molecular diffusion coefficient of the sludge, are the key parameters involved in the near field. The dilution effects of the far field migration patterns are also illustrated.

The geological research in France - The Dosssier 2005 Argile

2006 ◽  
Vol 985 ◽  
Author(s):  
Frederic Plas ◽  
Jacques WENDLING
Keyword(s):  
Near Field ◽  
Concept Design ◽  
Dose Calculations ◽  
Spent Fuel ◽  
Geological Medium ◽  
Source Data ◽  
Underground Research Laboratory ◽  
Broad Outline ◽  
Geological Research

AbstractAt the end of fifteen years of researchs defined by the French act of December 30, 1991 on radwaste management, Andra gave a report, “Dossier Argile 2005”, which concluded with the feasibility of a reversible disposal in the argillaceous Callovo-Oxfordien formation studied by means of an underground research laboratory at Meuse/Haute-Marne site. Starting from source data like the characteristics of the geological medium and the waste inventory, the process followed by Andra to achieve at this conclusion is of type sequential and iterative between concept design, scientific knowledge, in particular that of the phenomenological evolution of the reposiroty and its geological environment from operating period to long term, and Safety assessment. The “Dossier Argile 2005” covers a broad radwaste inventory, ILLW, HLW and Spent Fuel, so that it makes it possible to cover whole of the technological, scientific and safety topics. This article will give an overview of the geological disposal studies in France and draw the main conclusion of the Dossier 2005 Argile. It will be focused on the near field (Engineering components and near field host rock), while considering if necessary its integration within the whole system. After a short description of the concepts (incl. waste inventory and the characteristics of the Meuse/Haute the Marne site) and the functions of the components of repository and geological medium, one will describe successively the broad outline of the phenomenological evolution of repository and the geological medium in near field, by in particular releasing the time scales of processes and uncertainties of knowledge. On this basis, one will indicate the safety scenarios which were considered and the broad outline of performance and dose calculations. Lessons learn from the Dossier 2005 Argile will be discussed and perspective and priority for future will be indicated.

The Effect of Nearfield Constraints on the Corrosion Behavior of High Burnup Spent Fuel

2006 ◽  
Vol 932 ◽  
Author(s):  
Andreas Loida ◽  
Manfred Kelm ◽  
Bernhard Kienzler ◽  
Horst Geckeis ◽  
Andreas Bauer
Keyword(s):  
Host Rock ◽  
Near Field ◽  
Experimental Studies ◽  
Spent Fuel ◽  
Mutual Support ◽  
Significant Extent ◽  
Dissolution Rates ◽  
The Matrix ◽  
Container Material

ABSTRACTThe long-term immobilization for individual radioelements released from the waste form “spent fuel” in solid phases upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. Related experimental studies comprise effects induced by the presence of Fe based container material, and near field materials other than Fe for a rock salt environment. The effect of the presence of an argillaceous host rock containing organic matter and pyrite on fuel alteration was studied in addition. The results have shown that oxidative radio-lysis products were found to be consumed at a significant extent by the metallic Fe and by the argillaceous host rock. Under these conditions a decrease at a factor of ca.100 for both the matrix dissolution rates and the solution concentrations of U and Pu was found. There is mutual support between the matrix dissolution rates, the solution concentrations and the amounts of oxygen encountered during the experiments under various conditions controlled by the presence of near field materials under study.

Corrosion Behavior of High Burnup Spent Fuel in Highly Alkaline Solutions

2009 ◽  
Vol 1193 ◽  
Author(s):  
A. Loida ◽  
R. Gens ◽  
V. Metz ◽  
K. Lemmens ◽  
C. Cachoir ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Waste Product ◽  
Alkaline Solutions ◽  
Spent Fuel ◽  
Boom Clay ◽  
Matrix Bound ◽  
Engineered Barriers ◽  
Clay Formation ◽  
Fractional Release

AbstractThis study is focused on the alteration behavior of spent nuclear fuel when exposed to highly alkaline groundwater. Contact of highly alkaline solution with the waste product is considered in the Belgian concept for disposal in the Boom Clay formation. According to the “supercontainer design” the fuel will be encapsulated in carbon steel canisters, surrounded by a concrete over-pack. After saturation of the engineered barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Using this type of solution at pH 12.5, spent nuclear fuel corrosion experiments were conducted over 378 days. Under anoxic conditions, parallel experiments were performed (a) in the absence of Fe phases and (b) in the presence of solid Fe phases representing container (corrosion) products. Both types of experiments resulted in relatively low matrix dissolution rates, around 10-7 per day, according to the fractional release of Sr. Solution concentrations of actinides are close to or below the detection limit, indicating an effective retention of these radioelements in the system. The observed precipitation of a Ca rich phase onto the surfaces of the corroded fuel samples may be related to the inhibited re-lease of actinides, Sr and other matrix bound radioelements.

Safety Assessment of Geological Disposal of High-Level Radioactive Waste in Boom Clay: Relation With the Radionuclide Inventory

2009 ◽  
Author(s):  
Pierre Van Iseghem ◽  
Jan Marivoet
Keyword(s):  
Radioactive Waste ◽  
Safety Assessment ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
High Level ◽  
Clay Formation ◽  
Parameter Values

This paper discusses the impact of the parameter values used for the transport of radionuclides from high-level radioactive waste to the far-field on the long-term safety of a proposed geological disposal in the Boom Clay formation in Belgium. The methodology of the Safety Assessment is explained, and the results of the Safety Assessment for vitrified high-level waste and spent fuel are presented. The radionuclides having the strongest impact on the dose-to-man for both HLW glass and spent fuel are 79Se, 129I, 126Sn, 36Cl, and 99Tc. Some of them are volatile during the vitrification process, other radionuclides are activation products, and for many of them there is no accurate information on their inventory in the waste form. The hypotheses in the selection of the main parameter values are further discussed, together with the status of the R&D on one of the main dose contributing radionuclides (79Se).

The Second Safety Assessment and Feasibility Interim Report (SAFIR 2 Report) on HLW Disposal in Boom Clay: Overview of the Belgian Programme

2003 ◽  
Vol 807 ◽  
Author(s):  
Robert Gens ◽  
Philippe Lalieux ◽  
Peter De Preter ◽  
Ann Dierckx ◽  
Johan Bel ◽  
...  
Keyword(s):  
Safety Assessment ◽  
State Of The Art ◽  
Technical Feasibility ◽  
Geological Disposal ◽  
Boom Clay ◽  
Safety Case ◽  
Clay Formation ◽  
Main Barrier

ABSTRACTONDRAF/NIRAS – the Belgian radioactive waste management agency – has published in 2001 the SAFIR 2 report on request of the authorities. The SAFIR 2 report is to be considered as a state-of-the art report and not as a complete safety case. This report gives an overview of the Belgian R&D program related to the geological disposal of HLW and ILW for the period 1990–2000 in the Boom Clay (reference host rock). The three main outcomes of the SAFIR 2 report on which this paper will be more specifically focusing, are the following (including results reported after 2000): long-term safety functions, confirmation of the role of the Boom Clay formation as the main barrier and identification of practical difficulties with respect to technical feasibility (repository design).

Study of the Consequences of Secondary Water Radiolysis within and Surrounding a Defective Canister

2000 ◽  
Vol 663 ◽  
Author(s):  
Jinsong Liu ◽  
Bo Strömberg ◽  
Ivars Neretnieks
Keyword(s):  
Near Field ◽  
Experimental Studies ◽  
Solubility Limit ◽  
Balance Model ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Secondary Minerals ◽  
First Case ◽  
Bentonite Buffer

ABSTRACTA model has been developed to study the effects of secondary water radiolysis caused by dispersed radionuclides in a bentonite buffer surrounding a copper canister. The secondary radiolysis is the radiolysis caused by radionuclides that have been released from the spent fuel and are present either as solutes in the pore-water, as sorbed species on the surface of other minerals, or as secondary minerals. The canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The dissolution of the spent fuel is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases have been considered with the purpose to illustrate the behaviors of both conservative and non-conservative nuclides. The nuclides that are most relevant are those expected to be the dominant α-emitters in the long-term (e.g. 239Pu, 240Pu, 241Am). In the first case it is assumed that there is no precipitation of secondary minerals of the relevant radionuclides inside the canister. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The radionuclide released to the surrounding buffer is then predicted using a mass balance model. The modelling results show that in both cases, the spent fuel will not be oxidized at a rate significantly faster compared to the case where secondary radiolysis is completely neglected. In the first case, however, a large domain of the near-field can be oxidized due to a much faster depletion of reducing minerals in the buffer, compared to the case where secondary radiolysis is neglected. In the second case, the effects of the secondary water radiolysis will be quite limited.

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