Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts

Antti Poteri, Henrik Nordman, Veli-Matti Pulkkanen, Aimo Hautojärvi, Pekka Kekäläinen

Research output: Book/ReportBook (author)

Abstract

The repository system chosen in Finland for spent nuclear fuel is composed of multiple successive transport barriers. If a waste canister is leaking, this multi-barrier system retards and limits the release rates of radionuclides into the biosphere. Analysis of radionuclide migration in the previous performance assessments has largely been based on numerical modelling of the repository system. The simplified analytical approach introduced here provides a tool to analyse the performance of the whole system using simplified representations of the individual transport barriers. This approach is based on the main characteristics of the individual barriers and on the generic nature of the coupling between successive barriers. In the case of underground repository the mass transfer between successive transport barriers is strongly restricted by the interfaces between barriers leading to well-mixed conditions in these barriers. The approach here simplifies the barrier system so that it can be described with a very simple compartment model, where each barrier is represented by a single, or in the case of buffer, by not more than two compartments. This system of compartments could be solved in analogy with a radioactive decay chain. The model of well mixed compartments lends itself to a very descriptive way to represent and analyse the barrier system because the relative efficiency of the different barriers in hindering transport of solutes can be parameterised by the solutes half-times in the corresponding compartments. In a real repository system there will also be a delay between the start of the inflow and the start of the outflow from the barrier. This delay can be important for the release rates of the short lived and sorbing radionuclides, and it was also included in the simplified representation of the barrier system. In a geological multi-barrier system, spreading of the outflowing release pulse is often governed by the typical behaviour of one transport barrier, because the reservoir capacities of and mass transfer coefficients between adjacent barriers may differ significantly. Characterisation of these properties of the repository system by the simplified approach is straightforward. The relative efficiency of the different barriers in attenuating transport of radionuclides can be determined by comparing the solute's half-times in the barriers. Solute's half-times in different barriers can also be compared with the radioactive half-lives of the nuclides. Already the first barrier along the release path in which the solute's half-time is longer than the nuclide's radioactive half-life will be an efficient transport barrier for that nuclide, although the barrier with longest solute half-time will be the most efficient barrier. The release rates of radionuclides from a leaking waste canister may also be dominated by their source term instead of the barrier system of the repository. Spent nuclear fuel is a ceramic material that dissolves slowly into groundwater. Waste dissolution can also be treated as a barrier in which the dissolution time (or half of it) corresponds to a solute's half-times in a barrier, and can be readily compared with the other barriers. The validity of the simplified description was tested against numerical transport simulations for three representative nuclides: C-14, I-129 and Pu-239. The results of these simulations showed reasonable agreement with those of the simplified approach.
Original languageEnglish
PublisherPosiva
Number of pages92
ISBN (Print)978-951-652-201-5
Publication statusPublished - 2012
MoE publication typeC1 Separate scientific books

Publication series

SeriesPosiva-raportti - Posiva Report
Volume20
ISSN1239-3096

Fingerprint

disposal
solutes
radioactive isotopes
compartments
nuclides
cans
spent fuels
nuclear fuels
half life
mass transfer
dissolving

Keywords

  • radionuclide migration
  • repository
  • multibarrier system
  • performance assessment

Cite this

Poteri, A., Nordman, H., Pulkkanen, V-M., Hautojärvi, A., & Kekäläinen, P. (2012). Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts. Posiva . Posiva-raportti - Posiva Report, Vol.. 20
Poteri, Antti ; Nordman, Henrik ; Pulkkanen, Veli-Matti ; Hautojärvi, Aimo ; Kekäläinen, Pekka. / Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts. Posiva , 2012. 92 p. (Posiva-raportti - Posiva Report, Vol. 20).
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Poteri, A, Nordman, H, Pulkkanen, V-M, Hautojärvi, A & Kekäläinen, P 2012, Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts. Posiva-raportti - Posiva Report, vol. 20, Posiva .

Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts. / Poteri, Antti; Nordman, Henrik; Pulkkanen, Veli-Matti; Hautojärvi, Aimo; Kekäläinen, Pekka.

Posiva , 2012. 92 p. (Posiva-raportti - Posiva Report, Vol. 20).

Research output: Book/ReportBook (author)

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T1 - Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts

AU - Poteri, Antti

AU - Nordman, Henrik

AU - Pulkkanen, Veli-Matti

AU - Hautojärvi, Aimo

AU - Kekäläinen, Pekka

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N2 - The repository system chosen in Finland for spent nuclear fuel is composed of multiple successive transport barriers. If a waste canister is leaking, this multi-barrier system retards and limits the release rates of radionuclides into the biosphere. Analysis of radionuclide migration in the previous performance assessments has largely been based on numerical modelling of the repository system. The simplified analytical approach introduced here provides a tool to analyse the performance of the whole system using simplified representations of the individual transport barriers. This approach is based on the main characteristics of the individual barriers and on the generic nature of the coupling between successive barriers. In the case of underground repository the mass transfer between successive transport barriers is strongly restricted by the interfaces between barriers leading to well-mixed conditions in these barriers. The approach here simplifies the barrier system so that it can be described with a very simple compartment model, where each barrier is represented by a single, or in the case of buffer, by not more than two compartments. This system of compartments could be solved in analogy with a radioactive decay chain. The model of well mixed compartments lends itself to a very descriptive way to represent and analyse the barrier system because the relative efficiency of the different barriers in hindering transport of solutes can be parameterised by the solutes half-times in the corresponding compartments. In a real repository system there will also be a delay between the start of the inflow and the start of the outflow from the barrier. This delay can be important for the release rates of the short lived and sorbing radionuclides, and it was also included in the simplified representation of the barrier system. In a geological multi-barrier system, spreading of the outflowing release pulse is often governed by the typical behaviour of one transport barrier, because the reservoir capacities of and mass transfer coefficients between adjacent barriers may differ significantly. Characterisation of these properties of the repository system by the simplified approach is straightforward. The relative efficiency of the different barriers in attenuating transport of radionuclides can be determined by comparing the solute's half-times in the barriers. Solute's half-times in different barriers can also be compared with the radioactive half-lives of the nuclides. Already the first barrier along the release path in which the solute's half-time is longer than the nuclide's radioactive half-life will be an efficient transport barrier for that nuclide, although the barrier with longest solute half-time will be the most efficient barrier. The release rates of radionuclides from a leaking waste canister may also be dominated by their source term instead of the barrier system of the repository. Spent nuclear fuel is a ceramic material that dissolves slowly into groundwater. Waste dissolution can also be treated as a barrier in which the dissolution time (or half of it) corresponds to a solute's half-times in a barrier, and can be readily compared with the other barriers. The validity of the simplified description was tested against numerical transport simulations for three representative nuclides: C-14, I-129 and Pu-239. The results of these simulations showed reasonable agreement with those of the simplified approach.

AB - The repository system chosen in Finland for spent nuclear fuel is composed of multiple successive transport barriers. If a waste canister is leaking, this multi-barrier system retards and limits the release rates of radionuclides into the biosphere. Analysis of radionuclide migration in the previous performance assessments has largely been based on numerical modelling of the repository system. The simplified analytical approach introduced here provides a tool to analyse the performance of the whole system using simplified representations of the individual transport barriers. This approach is based on the main characteristics of the individual barriers and on the generic nature of the coupling between successive barriers. In the case of underground repository the mass transfer between successive transport barriers is strongly restricted by the interfaces between barriers leading to well-mixed conditions in these barriers. The approach here simplifies the barrier system so that it can be described with a very simple compartment model, where each barrier is represented by a single, or in the case of buffer, by not more than two compartments. This system of compartments could be solved in analogy with a radioactive decay chain. The model of well mixed compartments lends itself to a very descriptive way to represent and analyse the barrier system because the relative efficiency of the different barriers in hindering transport of solutes can be parameterised by the solutes half-times in the corresponding compartments. In a real repository system there will also be a delay between the start of the inflow and the start of the outflow from the barrier. This delay can be important for the release rates of the short lived and sorbing radionuclides, and it was also included in the simplified representation of the barrier system. In a geological multi-barrier system, spreading of the outflowing release pulse is often governed by the typical behaviour of one transport barrier, because the reservoir capacities of and mass transfer coefficients between adjacent barriers may differ significantly. Characterisation of these properties of the repository system by the simplified approach is straightforward. The relative efficiency of the different barriers in attenuating transport of radionuclides can be determined by comparing the solute's half-times in the barriers. Solute's half-times in different barriers can also be compared with the radioactive half-lives of the nuclides. Already the first barrier along the release path in which the solute's half-time is longer than the nuclide's radioactive half-life will be an efficient transport barrier for that nuclide, although the barrier with longest solute half-time will be the most efficient barrier. The release rates of radionuclides from a leaking waste canister may also be dominated by their source term instead of the barrier system of the repository. Spent nuclear fuel is a ceramic material that dissolves slowly into groundwater. Waste dissolution can also be treated as a barrier in which the dissolution time (or half of it) corresponds to a solute's half-times in a barrier, and can be readily compared with the other barriers. The validity of the simplified description was tested against numerical transport simulations for three representative nuclides: C-14, I-129 and Pu-239. The results of these simulations showed reasonable agreement with those of the simplified approach.

KW - radionuclide migration

KW - repository

KW - multibarrier system

KW - performance assessment

M3 - Book (author)

SN - 978-951-652-201-5

T3 - Posiva-raportti - Posiva Report

BT - Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts

PB - Posiva

ER -

Poteri A, Nordman H, Pulkkanen V-M, Hautojärvi A, Kekäläinen P. Representing Solute Transport Through the Multi-Barrier Disposal System by Simplified Concepts. Posiva , 2012. 92 p. (Posiva-raportti - Posiva Report, Vol. 20).