The electrochemical properties of alloy 690 in simulated pressurized water reactor primary water: Effect of temperature

Journal of Nuclear Materials - Tập 518 - Trang 305-315 - 2019
Jie Yang1, Yanhui Li1,2, Aoni Xu1,3, Balazs Fekete4, Digby D. Macdonald1
1Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720 USA
2Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi'an Jiaotong University, Xi'an, Shanxi, 710049, China
3Corrosion and Protection Center, Key Laboratory for Corrosion and Protection (MOE), University of Science and Technology Beijing, Beijing 100083, China
4Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA, 94720, USA

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Journal of Nuclear Materials

Tập 518

305-315

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