Thermal Behaviour of Backfill Material for a Nuclear Fuel Waste Disposal Vault

1989 ◽  
Vol 176 ◽  
Author(s):  
R.N. Yong ◽  
A.M.O. Mohamed ◽  
S.C.H. Cheung
Keyword(s):  
Mechanical Properties ◽  
Nuclear Fuel ◽  
Thermal Response ◽  
Management Program ◽  
Thermal And Mechanical Properties ◽  
Used Nuclear Fuel ◽  
Backfill Material ◽  
Rock Formations ◽  
Pressure Development ◽  
Nuclear Fuel Waste

ABSTRACTThe concept of disposing of used nuclear fuel in engineered rock formations is being studied in the Canadian Nuclear Fuel Waste Management Program. After the used fuel is emplaced in the vault, the vault would be backfilled. The backfill has to satisfy a number of engineering requirements. A reference backfill with satisfactory hydraulic, thermal and mechanical properties has already been selected.As the used fuel in the waste containers decays, heat will be generated and this heat will raise the temperature of the backfill material. The performance of the reference backfill material was evaluated over the temperature range 20–100°C. This paper addresses the results of experiments on thermal response and pressure development in the backfill for the period shortly after vault closure.

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.

The Canadian program for sealing underground nuclear fuel waste vaults

1984 ◽  
Vol 21 (3) ◽  
pp. 593-596 ◽  
Author(s):  
R. S. Lopez ◽  
S. C. H. Cheung ◽  
D. A. Dixon
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Chemical Species ◽  
Management Program ◽  
Crushed Rock ◽  
Compacted Clay ◽  
Rock Formations ◽  
Engineered Barriers ◽  
Shaft Seals ◽  
Nuclear Fuel Waste

The general objective of the Canadian Nuclear Fuel Waste Management Program is the safety assessment of the disposal of nuclear reactor wastes in underground vaults mined in plutonic rock formations of the Canadian Shield. The isolation of waste forms from the surface environment is effected by a system of multiple additive barriers, each one engineered to a high degree of excellence. In successive order, these barriers are the container, the buffer, the backfill of vault openings, shaft and drift seals, and plugs in boreholes.The buffer is primarily a chemical barrier designed to retard the migration of chemical species that may corrode the containers, and also the outward migration of radionuclides in case of container failure. The buffer is an engineered mixture of high-purity expandable clays and sand or crushed rock. The backfill material in drifts and shafts is primarily a physical barrier, designed to stabilize the rock excavations and to retard the movement of radionuclides by advection.Specially designed drift and shaft seals are being considered, to enhance the function of the backfill. The seals are engineered barriers that may include reinforced concrete bulkheads, compacted clay and sand plugs, and grouting of the surrounding rock. Key words: nuclear fuel waste, underground disposal, engineered barriers, swelling clay, buffer, backfill, shaft seals.

Xps and Xrd Studies of Uranium Rich Minerals from Cigar Lake, Saskatchewan

1987 ◽  
Vol 112 ◽  
Author(s):  
S. Sunder ◽  
P. Taylor ◽  
J. J. Cramer
Keyword(s):  
Nuclear Fuel ◽  
Photoelectron Spectroscopy ◽  
Uranium Concentration ◽  
Management Program ◽  
X Ray Diffraction ◽  
Xps Spectra ◽  
X Ray ◽  
Used Nuclear Fuel ◽  
Nuclear Fuel Waste ◽  
Lake Deposits

AbstractSamples of uranium ore from the 1,300 million year old Cigar Lake deposits in northern Saskatchewan have been investigated using X ray Photoelectron Spectroscopy (XPS) and X ray Diffraction (XRD). The XPS spectra showed the peaks for U, Pb, 0, C, Si, Ca and Mg. Analysis of highresolution XPS spectra showed that the oxidation state of uranium in these minerals is equal to, or lower than, that of U3O7. The XRD data confirm the presence of an oxidized phase, most likely α U3O7., in addition to UO2. The low U(VI)/U(IV) ratio in these minerals is consistent with the low uranium concentration found in groundwater in contact with these uranium minerals. The relevance of these results to the disposal concept of used nuclear fuel in the Canadian Nuclear Fuel Waste Management Program is discussed.

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.

Lifetimes of Titanium and Copper Containers for the Disposal of Used Nuclear Fuel

1991 ◽  
Vol 257 ◽  
Author(s):  
Lawrence H. Johnson ◽  
D.W. Shoesmith ◽  
B.M. Ikeda ◽  
F. King
Keyword(s):  
Nuclear Fuel ◽  
Bulk Solution ◽  
General Corrosion ◽  
Compacted Clay ◽  
Used Nuclear Fuel ◽  
Buffer Material ◽  
Using Data ◽  
Long Term Performance ◽  
Term Performance ◽  
Nuclear Fuel Waste

ABSTRACTTitanium and copper have been proposed as suitable container materials for disposal of nuclear fuel waste in plutonic rock of the Canadian Shield. Studies of the corrosion of these materials have led to the development of container failure models to predict long-term performance. Crevice corrosion and hydrogen-induced cracking of titanium have been identified as potential failure mechanisms, and these two processes have been studied in detail. Using data from these studies as well as a number of conservative assumptions, titanium container lifetimes of 1200 to 7000 a have been estimated. For copper, general corrosion has been studied in detail in bulk solution and in compacted clay-based buffer material. Results indicate that the copper corrosion rate is likely to be controlled by the rate of transport of copper species away from the container surface. An assessment of copper pitting data suggests that pitting is an extremely improbable failure mechanism. The copper container failure model predicts minimum container lifetimes of 30 000 a. The results demonstrate that long lifetime containment can be provided, should performance assessment studies indicate the need for such an option.

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.

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