Preliminary conceptual designs for advanced packages for the geologic disposal of spent fuel

ABSTRACTPursuant to the Decision-in-Principle of 2001 the Finnish programme for geologic disposal of spent fuel has now moved to the phase of underground characterisation of the repository site. The main objective of this programme phase is to confirm – or refute – the suitability of the Olkiluoto site by investigations conducted underground at the actual depth of the repository. The construction work of the access tunnel to the rock characterisation facility (ONKALO) started in the late summer of 2004. The site research and investigations work aims at the maturity needed for submission of the application for construction license of the actual repository in 2012. This requires, however, that also the technology has reached the maturity needed. The design and technical plans form the necessary platform for the development of the safety case for spent fuel disposal. A plan, “road map”, has been produced for the portfolio of reports that should demonstrate the safety of disposal as required by the criteria set by the government and further detailed by the safety authority, STUK.

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Inventories of Iodine-129 and Cesium-137 in the Gaps and Grain Boundaries of LWR Spent Fuels

MRS Proceedings ◽

10.1557/proc-556-487 ◽

1999 ◽

Vol 556 ◽

Cited By ~ 11

Author(s):

W. J. Gray

Keyword(s):

Grain Boundary ◽

Spent Nuclear Fuel ◽

Yucca Mountain ◽

Light Water Reactor ◽

Gas Release ◽

Spent Fuel ◽

Fission Gas ◽

Geologic Disposal ◽

Cesium 137 ◽

Fission Gas Release

AbstractPerformance assessment calculations that support geologic disposal of spent nuclear fuel in a potential repository at Yucca Mountain, Nevada, are based in part on the assumption that 2% of the total inventories of 135Cs, 129I, and 99Tc are located in the gap and grain-boundary regions where they could dissolve rapidly if the spent fuel were to be contacted by groundwater. Actual measured values reported here for a few light-water reactor (LWR) spent fuels show that the combined gap and grain-boundary inventories of 129I approximately equaled the fission-gas release fractions. For 137Cs, the combined gap and grain-boundary inventories were approximately one third of the fission-gas release fractions. These measured values can be used to replace the 2% estimate and thus reduce the uncertainties in the calculations.

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Materials Characteristics and Dissolution Behavior of Spent Nuclear Fuel

MRS Bulletin ◽

10.1557/s088376940004865x ◽

1994 ◽

Vol 19 (12) ◽

pp. 24-27 ◽

Cited By ~ 15

Author(s):

L.H. Johnson ◽

L.O. Werme

Keyword(s):

Nuclear Fuel ◽

Nuclear Waste ◽

Nuclear Power ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Waste Form ◽

Dissolution Behavior ◽

Spent Fuel ◽

Power Stations ◽

Geologic Disposal

The geologic disposal of spent nuclear fuel is currently under consideration in many countries. Most of this fuel is in the form of assemblies of zirconium-alloy-clad rods containing enriched (1–4% 235U) or natural (0.71% 235U) uranium oxide pellets. Approximately 135,000 Mg are presently in temporary storage facilities throughout the world in nations with commercial nuclear power stations.Safe geologic disposal of nuclear waste could be achieved using a combination of a natural barrier (the host rock of the repository) and engineered barriers, which would include a low-solubility waste form, long-lived containers, and clay- and cement-based barriers surrounding the waste containers and sealing the excavations.A requirement in evaluating the safety of disposal of nuclear waste is a knowledge of the kinetics and mechanism of dissolution of the waste form in groundwater and the solubility of the waste form constituents. In the case of spent nuclear fuel, this means developing an understanding of fuel microstructure, its impact on release of contained fission products, and the dissolution behavior of spent fuel and of UO2, the principal constituent of the fuel.

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Densification of Salt-Occluded Zeolite a Powders to a Leach-Resistant Monolith

MRS Proceedings ◽

10.1557/proc-333-277 ◽

1993 ◽

Vol 333 ◽

Cited By ~ 2

Author(s):

Michele A. Lewis ◽

Donald F. Fischer ◽

Christopher D. Murphy

Keyword(s):

Fast Reactor ◽

Fission Products ◽

Glass Frit ◽

Deionized Water ◽

Zeolite A ◽

Spent Fuel ◽

Release Rates ◽

Geologic Disposal ◽

Radiation Resistant ◽

Salt Waste

ABSTRACTPyrochemical processing of spent fuel from the Integral Fast Reactor (IFR) yields a salt waste of LiCI-KCI that contains approximately 6 wt% fission products, primarily as CsCI and SrCl2. Past work has shown that zeolite A will preferentially sorb cesium and strontium and will encapsulate the salt waste in a leach-resistant, radiation-resistant aluminosilicate matrix. However, a method is still needed to convert the salt-occluded zeolite powders into a monolith suitable for geologic disposal. We are thus investigating a method that forms bonded zeolite by hot pressing a mixture of glass frit and sait-occluded zeolite powders at 990 K (717°C) and 28 MPa. The leach resistance of the bonded zeolite was measured in static leach tests run for 28 days in 363 K (90°C) deionized water. Normalized release rates of all elements in the bonded zeolite were low, <1 g/m2d. Thus, the bonded zeolite may be a suitable waste form for IFR salt waste.

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