Effects of temperature, dissolved oxygen (DO), and degree of cold work (CW) on the oxidation kinetics of supercritical-water-cooled reactor (SCWR) fuel cladding candidate materials, including three types of 15Cr-20Ni austenitic stainless steels (1520 SSs), in superheated steam have been investigated assuming power-law kinetics. Characteristics of oxide layers and its relation to oxidation behaviors are also discussed. The effect of DO on the weight gain behavior in superheated steam at 700 °C was minor for all specimens at least up to 200 ppb DO. The tube-shaped specimens of 1520 SSs showed very good oxidation resistance at 700–780 °C. There was no clear difference in the oxidation kinetics among the three investigated types of 1520 SSs. The degree of CW is a significant parameter to mitigate oxidation in superheated steam. It has been suggested that the tube specimens showed a very slow oxidation kinetics since Cr diffusion in the outside surface of the tubes is accelerated as a result of an increase of dislocation density and/or grain refinement by a high degree of CW.
Skip Nav Destination
2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5578-2
PROCEEDINGS PAPER
High-Temperature Steam Oxidation Kinetics and Mechanism of SCWR Fuel Cladding Candidate Materials
Seung Mo Hong,
Seung Mo Hong
Tohoku University, Sendai, Japan
Search for other works by this author on:
Yutaka Watanabe
Yutaka Watanabe
Tohoku University, Sendai, Japan
Search for other works by this author on:
Hiroshi Abe
Tohoku University, Sendai, Japan
Seung Mo Hong
Tohoku University, Sendai, Japan
Yutaka Watanabe
Tohoku University, Sendai, Japan
Paper No:
ICONE21-16432, V001T02A038; 6 pages
Published Online:
February 7, 2014
Citation
Abe, H, Hong, SM, & Watanabe, Y. "High-Temperature Steam Oxidation Kinetics and Mechanism of SCWR Fuel Cladding Candidate Materials." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications. Chengdu, China. July 29–August 2, 2013. V001T02A038. ASME. https://doi.org/10.1115/ICONE21-16432
Download citation file:
9
Views
Related Proceedings Papers
Related Articles
Microstructure Study of NiCrAlY and FeCrAlY Exposed to Superheated Steam at 800 °C
ASME J of Nuclear Rad Sci (January,2018)
Assessment of Candidate Fuel Cladding Alloys for the Canadian Supercritical Water-Cooled Reactor Concept
ASME J of Nuclear Rad Sci (January,2016)
European Project “Supercritical Water Reactor-Fuel Qualification Test”: Summary of General Corrosion Tests
ASME J of Nuclear Rad Sci (July,2016)
Related Chapters
E110opt Fuel Cladding Corrosion under PWR Conditions
Zirconium in the Nuclear Industry: 20th International Symposium
Combined Cycle Power Plant
Energy and Power Generation Handbook: Established and Emerging Technologies
Effect of Chromium Content on the On-Cooling Phase Transformations and Induced Prior-β Zr Mechanical Hardening and Failure Mode (in Relation to Enhanced Accident-Tolerant Fuel Chromium-Coated Zirconium-Based Cladding Behavior upon and after High-Temperature Transients)
Zirconium in the Nuclear Industry: 20th International Symposium