Long-Term Thermal Stability of Alloy 825 as a High-Level Nuclear Waste Container Material

1995 ◽  
Vol 412 ◽  
Author(s):  
D. S. Dunn ◽  
Y.-M. Pan ◽  
G. A. Cragnolino ◽  
N. Sridhar
Keyword(s):  
Grain Boundary ◽  
Nuclear Waste ◽  
Elevated Temperatures ◽  
Thermal Exposure ◽  
Boundary Region ◽  
Waste Container ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Welding Processes ◽  
Container Material

AbstractThe thermal exposure of Fe-Cr-Ni-Mo materials to certain temperature regimes often results in the formation of grain boundary carbides and the associated depletion of alloying elements. This phenomenon, termed sensitization, is frequently the result of welding processes or in service exposure to elevated temperatures. In this investigation, alloy 825, a candidate high-level nuclear waste (HLW) container material, was thermally exposed to temperatures in the range of 550 to 800 °C for periods of up to 1,000 hr. Sensitization of the material was evaluated by corrosion tests and grain boundary analyses using an analytical electron microscope. The sensitized microstructure was found to contain M23C6-type carbides as well as a Cr-depleted region in the vicinity of the grain boundaries. The degree of sensitization was correlated to the extent of Cr depletion in the grain boundary region.

Evaluation of the Susceptibility of Astm A216 Grade Wca Mild Steel to Stress Corrosion Cracking in Simulated Salt Repositopy Environments

1986 ◽  
Vol 84 ◽  
Author(s):  
S. G. Pitman
Keyword(s):  
Mild Steel ◽  
Stress Corrosion ◽  
Nuclear Waste ◽  
Dynamic Tests ◽  
Static Tests ◽  
Waste Package ◽  
Test Conditions ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Container Material

AbstractIn current conceptual designs, a mild steel (ASTM A?16 Grade WCA) is the relerence container material for use in high level nuclear waste packages intended for emplacement in a salt repository. The resistance of the steel to stress corrosion crackinq (SCC) is being investigated as part of the effort underway to verify the suitability of the material for waste package applications. Static tests (U-bend and bolt-loaded fracture toughness specimens) and dynamic tests (slow strain rate and corrosion fatigue) were conducted on both as-cast and weldment specimens of the material, in both low-Mg and high-Mg halite-saturated brines, in the temperature range of 90 to 200°C. The investigations indicate that the steel is not susceptible to SCC under the test conditions employed.

The Interaction of Na-Al-P-Glass (Cs, Sr-Bearing) with Water at Elevated Temperatures (70–250°C)

1994 ◽  
Vol 353 ◽  
Author(s):  
Alexander P. Mukhamet-Galeyev ◽  
Larisa O. Magazina ◽  
Konstantin A. Levin ◽  
Nikolay D. Samotoin ◽  
Alexander V. Zotov ◽  
...  
Keyword(s):  
Nuclear Waste ◽  
Elevated Temperatures ◽  
Solid Phase ◽  
Geochemical Behavior ◽  
Slowing Down ◽  
Glass Dissolution ◽  
Detailed Composition ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Glass Leaching

At present Na-Al-P-glass is used in Russia as matrix for high level nuclear waste which include a sufficient portion of 90Sr and 137Cs isotopes. It has been known that this type of glass is characterized by relatively poor hydrothermal stability above 120°C (Krilova et al., 1990 [1], Balukova et al., 1987 [2]). Although careful studies were devoted to the glass, some questions of geochemical behavior of Sr and Cs in the limits of the nearest repository environment have remained undecided. For example, these have been left uncertain: the mechanism of the glass leaching, phenomena limiting the glass dissolution, the detailed composition of newly formed phases, temperature dependence of the glass solubility. The main purpose of the present experimental investigation was to detect factors controlling Sr and Cs concentrations of aquous solutions and to study the possibility of new solid phase formation as a factor slowing down migration of Cs and Sr in water satureted circumstance.

scholarly journals Container Materials for High-Level Nuclear Waste at the Proposed Yucca Mountain Site

1990 ◽  
Vol 212 ◽  
Author(s):  
R. Daniel McCright
Keyword(s):  
Nuclear Waste ◽  
Yucca Mountain ◽  
Localized Corrosion ◽  
Experimental Program ◽  
Degradation Rates ◽  
Design Activities ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Welding Processes

ABSTRACTCandidate container materials for high-level nuclear waste packages to be emplaced at the proposed Yucca Mountain repository site are being considered for their long-term resistance to corrosion, oxidation, embrittlement and other kinds of degradation. Selection criteria have been established, and a method has been developed for recommending a material for advanced container design activities. An extensive compilation of the degradation phenomena for six candidate materials is complete, and further studies have begun on the degradation modes affecting additional candidate materials. Phenomenological models for predicting container degradation rates are being advanced for environmental conditions applicable to Yucca Mountain. An experimental program is underway to evaluate the susceptibility of container materials to localized corrosion, stress corrosion cracking, and enhancement of corrosion and oxidation attack by gamma radiation. Initial evaluations of container fabrication and welding processes have identified some processes that appear to alleviate some long-term corrosion susceptibility concerns.

Transmutation of high-level nuclear waste by means of accelerator driven system

2013 ◽  
Vol 3 (1) ◽  
pp. 60-69 ◽  
Author(s):  
Hamid Aït Abderrahim ◽  
Didier De Bruyn ◽  
Gert Van den Eynde ◽  
Sidney Michiels
Keyword(s):  
Nuclear Waste ◽  
Driven System ◽  
High Level Nuclear Waste ◽  
Accelerator Driven System ◽  
High Level
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