Metal Matrix Integrity and Related Technology Development in the Canadian Nuclear Fuel Waste Management Program

1983 ◽  
Vol 26 ◽  
Author(s):  
P. Mani Mathew ◽  
Paul A. Krueger
Keyword(s):  
Nuclear Fuel ◽  
Metal Matrix ◽  
Technology Development ◽  
Management Program ◽  
Matrix Cracking ◽  
Thin Wall ◽  
Controlled Cooling ◽  
Casting System ◽  
Metallurgical Characterization ◽  
Nuclear Fuel Waste

ABSTRACTOne of the concepts under development as a nuclear fuel waste isolation container is a thin-wall corrosion-resistant shell, supported internally by a cast metal matrix in which intact used fuel bundles are invested. The integrity of the metal matrix can be influenced by metallurgical factors and by process parameters. Finite element solidification modelling and laboratory experiments with lead as an investment material have shown the influence of heat transfer parameters on matrix integrity. Controlled cooling of the container walls, for example, can be used to reduce the interaction time between the molten matrix, the fuel sheathing and the container wall, and achieve a void-free matrix. The results of the computer simulations have been used to design an improved casting system, based on controlled wall cooling, for investing nuclear fuel waste containers. Ultrasonic testing of bonds between some candidate container and metal matrix materials, in combination with the metallurgical characterization of the interface region, has allowed differentiation between bonded and unbonded regions. Matrix cracking near bonded interfaces was identified as a potential problem, which could limit the use of the ultrasonic scanning technique for matrix inspection. To produce a high quality interface with good chemical bonding, induction skin melting looks promising and is being further evaluated.

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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