Review of Corrosion Modes For Alloy 22 Regarding Lifetime Expectancy of Nuclear Waste Containers
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Review of Corrosion Modes For Alloy 22 Regarding Lifetime Expectancy of Nuclear Waste Containers Raúl B. Rebak and John C. Estill Lawrence Livermore National Laboratory 7000 East Ave. L-631 Livermore, CA 94550, USA
ABSTRACT Alloy 22 (UNS N06022) was selected to fabricate the corrosion resistant outer barrier of a two-layer nuclear waste package container. This paper reviews the main corrosion degradation modes that are predicted for the outer layer of the container. Current results show that the containers would perform well under general corrosion, localized corrosion and environmentally assisted cracking (EAC). For example, the general corrosion rate is expected to be below 100 nm/year and the container is predicted to be outside the range of potential for localized corrosion and environmentally assisted cracking.
INTRODUCTION Yucca Mountain (Nevada) is being characterized as the site for geological disposal of commercial nuclear spent fuel and some other forms of high-level nuclear waste in the United States. Yucca Mountain is located about 160 km northwest of Las Vegas in the state of Nevada on land owned by the federal government [1]. The overall strategy in isolating highlevel nuclear waste is to make use of the natural barriers present in the host geologic site along with the construction of a series of engineered barriers. The current waste package design consists of two concentric metal containers. The outer container would be made of Alloy 22 or UNS N06022, which is among the most corrosion resistant of all engineering materials. The approximate composition of Alloy 22 (in weight %) is: ~56 Ni, 22 Cr, 13 Mo, 3 W and 3 Fe. The purpose of this outer container is to provide protection against corrosion. The inner container would be thicker and made of nuclear grade type 316L stainless steel or UNS S31603. The intended purpose of the inner barrier is to provide shield for radiation and mechanical integrity. The engineering barrier design also includes a detached drip shield that would be emplaced above the waste package to deflect any falling water or rocks from the walls of the emplacement onto the containers. The proposed material for the drip shield is Titanium Grade 7 or UNS R52400 [1]. It is required that the containers would not release radioactive material to the surrounding mountain for several thousands of years. Due to radioactive decay, the temperature of the container may rise to a maximum of nearly 160°C during the first ~1,000 years. The magnitude of the temperature depends on the number of containers per unit length of the emplacement tunnels. The repository will be located above the local water table. However, aqueous solutions may enter in contact with the container surfaces through two different mechanisms, namely (1) Dripping from the drift wall above the container (in absence of the drip shield) and (2) At a given relative humidity, through the deliquescence of salt or mountain (mineral) dust that may accumulate on top of the container during drier periods. In both cases the aque
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