Immobilization and geological disposal of nuclear fuel waste

1984 ◽  
Vol 62 (8) ◽  
pp. 979-985
Author(s):  
J. C. Tait
Keyword(s):  
Nuclear Fuel ◽  
Glass Ceramics ◽  
Management Program ◽  
Crystalline Materials ◽  
Dissolution Behaviour ◽  
Geological Formation ◽  
High Level ◽  
Radionuclide Release ◽  
Interim Storage ◽  
Nuclear Fuel Waste

The Canadian Nuclear Fuel Waste Management Program is developing methods for the safe disposal of both used nuclear fuel and fuel recycle waste. The disposal strategy is based on interim storage of the used fuel, immobilization of either used fuel or recycle waste, and disposal, deep in a stable geological formation in the Canadian Shield. The disposal concept proposes a multibarrier system to inhibit the release of the radioactive waste from the disposal vault. The principal components of the multibarrier system are (i) the waste form in whch the radionuclides are immobilized, (ii) engineered barriers including high integrity containers, buffers and backfills designed to retard the movement of groundwaters in the disposal vault, and (iii) the natural barrier provided by the massive geological formation itself. The research programs to investigate this concept are discussed briefly. Several different waste forms are being developed for the immobilization of high-level fuel recycle waste, including glass, glass-ceramics and crystalline materials. Dissolution of these materials in groundwater is the only likely scenario that could lead to radionuclide release. The factors that influence the aqueous dissolution behaviour of these materials are reviewed.

Key Aspects of the H12 Safety Case

2000 ◽  
Vol 663 ◽  
Author(s):  
Hiroyuki Umeki
Keyword(s):  
Nuclear Fuel ◽  
General Public ◽  
Spent Nuclear Fuel ◽  
Atomic Energy Commission ◽  
Management Program ◽  
Geological Disposal ◽  
Safety Case ◽  
Geological Formation ◽  
High Level ◽  
Key Aspects

ABSTRACTIn Japan, as outlined in the overall high-level radioactive waste (HLW) management program defined by the Japanese Atomic Energy Commission (AEC, 1994), HLW from reprocessing of spent nuclear fuel will be immobilized in a glass matrix and stored for a period of 30 to 50 years to allow cooling. It will then be disposed of in a deep geological formation. Pursuant to the overall HLW management program, an organization with responsibility for implementing HLW disposal will be established around the year 2000. This will be followed by site selection and characterization, demonstration of disposal technology, establishment of the necessary legal infrastructure, relevant licensing applications and repository construction, with the objective of starting repository operation by the 2030s and no later than the mid 2040s.The HLW disposal program is currently in the research and development (R&D) phase and the Japan Nuclear Cycle Development Institute (JNC) has been assigned as the leading organization responsible for R&D activities. The aim of the R&D activities at the current stage is to provide a scientific and technical basis for the geological disposal of HLW in Japan and to promote understanding of the safety concept not only in the scientific and technical community but also by the general public. One of the features of the R&D program is that its progress is documented at appropriate intervals, with a view to clearly determining the level of achievement of the program and to promote understanding and acceptance of the geological disposal strategy by the general public. As a major milestone, the Power Reactor and Nuclear Fuel Development Corporation (PNC, now JNC) submitted a first progress report, referred to as H3 (PNC, 1992), in September 1992.

Materials Research for the Canadian Nuclear Fuel Waste Management Program

1981 ◽  
Vol 6 ◽  
Author(s):  
Donald J. Cameron
Keyword(s):  
Nuclear Fuel ◽  
Waste Disposal ◽  
Hard Rock ◽  
Technology Development ◽  
Management Program ◽  
Materials Research ◽  
Backfill Materials ◽  
Nuclear Fuel Waste ◽  
Irradiated Fuel ◽  
Selection Of

ABSTRACTNuclear fuel waste disposal research in Canada is concentrating on hard-rock disposal. The research programs concerned with the man-made components of the disposal system are reviewed. Irradiated fuel and solidified reprocessing wastes are both being researched, as are durable containers, and buffer and backfill materials. This review concentrates mainly on the more scientific aspects of the research, which contribute to the selection of preferred options for the various components of the system, and which support directly or indirectly the safety analysis of the disposal concept. Some technology development is included in the program now, and this is expected to expand as confidence in the acceptability of the disposal concept grows.

A Review of Progress in the Canadian Nuclear Fuel Waste Management Program

1985 ◽  
Vol 50 ◽  
Author(s):  
R. B. Lyon ◽  
L. H. Johnson
Keyword(s):  
Waste Management ◽  
Nuclear Fuel ◽  
Management Program ◽  
Selection Methods ◽  
Transportation Fuel ◽  
Waste Immobilization ◽  
Fuel Storage ◽  
And Performance ◽  
Nuclear Fuel Waste ◽  
Made In

AbstractThe Canadian Nuclear Fuel Waste Management Program is reviewed, illustrating the progress that has been made in assessing the concept of disposal of nuclear fuel waste in plutonic rock of the Canadian Shield. Research is being conducted into used fuel storage and transportation, fuel waste immobilization, site characterization and selection methods, and performance assessment modelling. Details of achievements in these areas are outlined, and results of the most recent interim assessment are discussed.

Predicted Behavior of Technetium in a Geological Disposal Vault For Used Nuclear Fuel - Ramifications of a Recent Determination of the Enthalpy of Formation of TcO2(cr)

1995 ◽  
Vol 412 ◽  
Author(s):  
R. J. Lemire ◽  
D. J. Jobe
Keyword(s):  
Nuclear Fuel ◽  
Management Program ◽  
Heat Of Solution ◽  
Used Nuclear Fuel ◽  
A Value ◽  
Reducing Conditions ◽  
Congruent Dissolution ◽  
Nuclear Fuel Waste ◽  
The Enthalpy Of Formation

AbstractRecently, we reported a value of ΔH°(TcO2(cr)) = -(458 ± 6) kJ·mol-1based on heat of solution measurements. The implications of this value on the database used in the Canadian Nuclear Fuel Waste Management Program for the evaluation of the technetium released by congruent dissolution of used UO2 fuel have now been assessed.It is probable that the Tc(IV) oxides are more stable than previously predicted and, hence, they are less likely to be oxidized to TcO4(aq) under moderately reducing conditions. We have revised earlier calculations done to predict the solution concentrations of technetium species in a vault as a function of the oxidation conditions in model groundwaters.

Prediction of the Oxidative Dissolution Rates of Used Nuclear Fuel in a Geological Disposal Vault Due to the Alpha Radiolysis of Water

1994 ◽  
Vol 353 ◽  
Author(s):  
S. Sunder ◽  
D.W. Shoesmith ◽  
N.H. Miller
Keyword(s):  
Nuclear Fuel ◽  
Management Program ◽  
Oxidative Dissolution ◽  
Dissolution Rates ◽  
Used Nuclear Fuel ◽  
A Value ◽  
Thin Layer Cell ◽  
Alpha Radiolysis ◽  
Nuclear Fuel Waste ◽  
Electrochemical Model

AbstractEffects of alpha radiolysis of water on the corrosion of nuclear fuel (UO2) have been investigated in solutions at pH = 9.5, i.e., a value close to that expected in groundwaters at the depth of the disposal vault proposed in the Canadian nuclear fuel waste management program, CNFWMP. The corrosion potentials of UO2 electrodes exposed to the products of alpha radiolysis of water were monitored as a function of alpha flux and exposure time in a specially designed thin-layer cell. The oxidative dissolution rates of UO2 are calculated from the steady-state values of the corrosion potential using an electrochemical model. A procedure to predict the dissolution rate of used nuclear fuel in groundwater as a function of fuel cooling time is described, and illustrated by calculating the dissolution rates for the reference used fuel in the CNFWMP (Bruce CANDU reactor fuel, burnup 685 GJ/kg U). It is shown that the oxidative dissolution of used fuel in the CNFWMP will be important only for time periods ≤ 600 a at this burnup and assuming no decrease in pH.

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