B. M. Ikeda

Hydrogen Absorption and the Lifetime Performance of Titanium Nuclear Waste Containers

For the conditions expected in a Canadian nuclear waste vault, the two corrosion processes most likely to lead to the failure of titanium waste containers are crevice corrosion and/or hydrogen induced cracking (HIC). In this report the processes likely to lead to hydrogen absorption by titanium alloys (Grades-2, -12, -16), and hence to render them susceptible to HIC are discussed. The possible paths to container failure via a combination of crevice corrosion, general passive corrosion and HIC are described and a criterion for container failure by HIC is defined. The modelling procedures developed to predict the consequences of hydrogen absorption are described, and the experimental methods used to measure required modelling parameters are discussed and the results summarized. The predictions of these models show that for alloys on which crevice corrosion either does not occur or is limited in extent (Ti-16 and Ti-12), container lifetimes of >10 a are achievable. As a result only those containers assumed to fail prematurely due to the presence of manufacturing defects need be included in assessment calculations.

Hydrogen-induced cracking of commercial purity titanium

Slow straining of compact tension specimens of commercial-purity titanium has been employed for assessing the likelihood of failure due to hydrogen pick-up in titanium containers for nuclear fuel waste disposal. Results indicate that slow crack growth occurs by a ductile tearing process at low hydrogen concentrations. No fast crack growth occurs at such hydrogen levels, apparently because ductile collapse relaxes the stresses and prevents the attainment of a sufficiently high stress intensity for fast crack initiation. Fast propagation of a brittle crack was observed only at hydrogen concentrations above a critical value that depended upon the material involved and the orientation of the crack relative to the manufactured microstructure. An empirical relationship suggests that the higher the strength of the titanium involved the lower is the critical hydrogen level for brittle failure. Both the distribution of residual β-phase and the texture of the fabricated material influence the susceptibility of a particular specimen orientation to fast fracture.

Modeling the Failure of Nuclear Waste Containers

Abstract A model was developed to predict the failure of Grade-2 titanium (Ti-2) nuclear waste containers. Two major corrosion modes were included: failure by crevice corrosion (CC) and failure by hydrogen-induced cracking (HIC). A small number of containers were assumed to be defective and to fail within 50 years of emplacement. The model is probabilistic in nature, and each modeling parameter was assigned a range of values, resulting in a distribution of corrosion rates and failure times. The crevice corrosion rate (RCC) was assumed to be dependent only upon properties of the material used and the temperature of the vault. CC was assumed to initiate rapidly on all containers and to propagate indefinitely without repassivation. Failure by HIC was assumed to be inevitable once container temperature (T) fell to ≤ 30°C. Depending upon the rate at which they were expected to cool, temperature-time profiles for individual containers were approximated by two-step or single-step temperature-time functions. Thes...

Modelling Procedures for Predicting the Lifetimes of Nuclear Waste Containers

Approaches to modelling the corrosion of nuclear waste containers are reviewed. The required containment of many thousands of years makes this a daunting task. The process has been simplified for a disposal vault in which redox conditions evolve from initially oxidizing to eventually non-oxidizing. The corrosion behaviour can be divided into two periods: an early hot, oxidizing period when localized corrosion damage is to be expected; and a later cool, non-oxidizing period when localized processes would be stifled, or repassivated, and general corrosion will predominate. At present, deterministic models to predict localized corrosion damage during the early period are unavailable or, at best, preliminary. Generally, the approach to predicting localized penetration of the container has been stochastic in nature and extreme value statistical analyses have been used to predict the expected penetration of carbon steel or copper containers by pitting. Experiments to determine the rate of crevice propagation in titanium are discussed and a model developed to predict failure of titanium waste containers by either crevice corrosion or hydrogen-induced cracking described. General corrosion occurring in the second, less oxidizing, period is more amenable to modelling by deterministic methods. Models based on electrochemical descriptions of the interfacial kinetics are described for carbon steel and copper containers, two materials expected to corrode actively under waste vault conditions. To date, no adequate model exists to predict the slow general corrosion of passivated materials.

The Stress Corrosion Cracking of Copper:Nuclear Waste Containers

The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a “limited propagation” argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J ox ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NaNO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J ox are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained.

Effect of Radiation on the Corrosion of Candidate Materials for Nuclear Waste Containers

The published information on the effect of radiation on the corrosion of potential waste container materials is reviewed. The materials discussed are irons and carbon steels, copper and copper alloys, stainless steels, nickel-based alloys, and titanium alloys. In general, a dose rate of >5 Gy·h -1 is required to accelerate the general corrosion of corrosionallowance materials. The effects of radiation on the localized corrosion of corrosion-resistant materials appear to be contradictory. In some cases, localized corrosion is induced, whereas in others it is inhibited.

Investigation of Thermophysical and Nuclear Properties of Prospective Coolants for Generation-IV Nuclear Reactors

Recently, a group of countries has initiated an international collaboration, the Generation IV International Forum (GIF), to develop the next-generation nuclear reactors. The GIF program has narrowed the design options of nuclear reactors to the following six concepts:1) SuperCritical-Water-cooled Reactor (SCWR);2) Sodium-cooled Fast Reactor (SFR);3) Lead-cooled Fast Reactor (LFR);4) Molten Salt Reactor (MSR);5) Gas-cooled Fast Reactor (GFR); and6) Very-High-Temperature Reactor (VHTR);The purpose of this paper is to compare main thermophysical, corrosion, and neutronic properties of the Generation-IV reactors’ coolants within the proposed range of operation.Copyright

Study on Primary and Secondary Heat-Transport Systems for Sodium-Cooled Fast Reactor

One of the current engineering challenges is to design next generation (Generation IV) Nuclear Power Plants (NPPs) with significantly higher thermal efficiencies (43–55%) compared to those of current NPPs to match or at least to be close to the thermal efficiencies reached at fossil-fired power plants (55–62%). The Sodium-cooled Fast Reactor (SFR) is one of the six concepts considered under the Generation IV International Forum (GIF) initiative.The BN-600 reactor is a sodium-cooled fast-breeder reactor built at the Beloyarsk NPP in Russia. This concept is the only one from the Generation IV nuclear-power reactors, which is actually in operation (since 1980’s). At the secondary side, it uses a subcritical-pressure Rankine-steam cycle with heat regeneration. The reactor generates electrical power in the amount of 600 MWel. The reactor core dimensions are 0.75 m (height) by 2.06 m (diameter). The UO2 fuel enriched to 17–26% is utilized in the core.There are 2 loops (circuits) for sodium flow. For safety reasons, sodium is used both in the primary and the intermediate circuits. Therefore, a sodium-to-sodium heat exchanger is used to transfer heat from the primary loop to the intermediate one. In this work major parameters of the reactor are listed. The actual scheme of the power-conversion heat-transport system is presented; and the results of the calculation of thermal efficiency of this scheme are analyzed. Details of the heat-transport system, including parameters of the sodium-to-sodium heat exchanger and main coolant pump, are presented.In this paper two possibilities for the SFR in terms of the power-conversion cycle are investigated: 1. a subcritical-pressure Rankine-steam cycle through a heat exchanger (current approach in Russian and Japanese power reactors); 2. a supercritical-pressure CO2 Brayton gas-turbine cycle through a heat exchanger (US approach).With the advent of modern super-alloys, the Rankine-steam cycle has progressed into the supercritical region of the coolant and is generating thermal efficiencies into the mid 50% range. Therefore, the thermal efficiency of a supercritical Rankine-steam cycle is also briefly discussed in this paper.According to GIF, the Brayton gas-turbine cycle is under consideration for future nuclear power reactors. The supercritical-CO2 cycle is a new approach in the Brayton gas-turbine cycle. Therefore, dependence of the thermal efficiency of this SC CO2 cycle on inlet parameters of the gas turbine is also investigated.Copyright

Model Passivated Carbon Electrodes for Fluorine Generation in MSRs and the Nuclear Fuel Cycle

Elemental fluorine, F2, is used in the nuclear fuel cycle for the isotopic separation of uranium-235 and 238, as well as for the purification of LiF-BeF2 in molten salt reactors. F2 is generated on an industrial scale by an electrochemical process using carbon electrodes in a KF-2HF molten salt. Carbon electrodes are used for industrial F2 generation due to its chemical stability, high conductivity, and relatively low cost. One of the main issues faced when using carbon electrodes in this chemical system is passivation through the formation of C-F compounds on the surface of the electrode. This results in a loss of anode wettability to the electrolyte and diminished charge transfer rate. The voltage needed for the fluorine evolution reaction increases which negatively impacts the safety of the system, increases the operating costs, and leads to faster degradation of the electrode.The degradation of electrical properties during passivation is progressive, eventually leading to electrode deactivation. The process of deactivation begins with a passivating C-F layer at potentials above the equilibrium potential (2.92 V). The layer is both non-wetting to the KF-2HF media and insulating. Deactivation begins with inhibited F2 bubble detachment, formation of a persistent gas layer, and finally deactivation as the electrode surface is completely covered by a thick, insulating C-F layer causing charge transfer to cease. Only a small current is able to flow, even at high potentials (up to 9 V), indicating F2 generation is completely inhibited. The purpose of this study is to manufacture and test model carbon electrodes and, to examine the non-wetting properties of a partially fluorinated surface. The electrodes will be prepared by mixing PTFE-particles with Vulcan carbon powder and then pressing to form pellets. These electrodes should have a reproducible surface for electrochemical performance studies that will lead to a better understanding of the surface chemistry. The research will develop novel electrodes with a goal to minimize the voltage required for F2 production. This will enhance the efficiency in the overall process and lower the manufacturing costs for F2.Carbon electrodes with different PTFE-content (20 w.% and 35 w.%) were synthesized. Electrochemical fluorination was then carried out at different potentials in the F2 generation region (4 to 8 V) in molten KF·2HF electrolyte at ∼90 °C. The electrochemical behaviour of the carbon-PTFE electrodes was examined and compared for both fluorine passivated and non-passivated graphite, amorphous carbon, and vitreous carbon electrodes. The growth of the electrical double-layer capacitance between the carbon electrodes and the KF·2HF molten salt was studied. The effects of composition of fluorinated and non-fluorinated carbon on electrode performance are presented.© 2013 ASME

Centralization of Canada’s Spent Nuclear Fuel

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel.The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method.The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.Copyright