The anodic behaviour of low molybdenum-uranium alloys at higher electrolyte temperatures

Mance, A.

doi:10.1016/0022-3115(69)90221-9

Cited by 5 publications

(2 citation statements)

References 4 publications

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“…accelerates the process of anodic dissolution. (Mance 1969) (Key words: corrosion, electrochemistry, molybdenum, reaction rate, uranium alloys) "Uranium-Molybdenum Alloys in Reactor Construction" This is a good general reference on U-Mo alloys. Information about fabrication, phase diagram, mechanical properties, radiation resistance, corrosion resistance are included.…”

Section: General References On Corrosion (5)mentioning

confidence: 99%

“…(Kim, Kim et al 1999) "The anodic behavior of low molybdenum-uranium alloys at higher electrolyte temperatures" See Summary above. (Mance 1969) "An Analytical Study of Gas-Bubble Nucleation Mechanisms in Uranium-Alloy Nuclear Fuel at High Temperature" A multi-atom gas-bubble nucleation mechanism in uranium-alloy nuclear fuel operating in the high-temperature equilibrium gamma phase is proposed based on interpretation of measured intragranular bubble-size distribution data. An aanalytical equation model is presented for the fission bubble nucleation, growth and coalescence.…”

Section: U-mo (8)mentioning

confidence: 99%

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Corrosion report for the U-Mo fuel concept

Henager

Bennett

Doherty

et al. 2014

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The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30˚C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

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Section: General References On Corrosion (5)mentioning

confidence: 99%

Section: U-mo (8)mentioning

confidence: 99%