Development of alumina-forming duplex stainless steels as accident-tolerant fuel cladding materials for light water reactors

Hyunmyung Kim, Hun Jang, Gokul Obulan Subramanian, Chaewon Kim, Changheui Jang*

*Corresponding author for this work

Research output: Contribution to journalArticlepeer-review

16 Citations (Scopus)


High aluminum content (>5 wt.%) duplex stainless steels were developed as accident-tolerant fuel (ATF) cladding materials for light water reactors. A set of model alumina-forming duplex stainless steel (ADSS) alloys with a nominal composition range of Fe-(18–21)Ni-(16–21)Cr-(5–6)Al were prepared and tested for corrosion resistance in both 1200 °C steam and simulated pressurized water reactor (PWR) operating conditions. The performance and properties of the ADSS alloys were compared to those of the austenitic (310 S) and ferritic (FeCrAl-APM) stainless steels. The results showed that the ADSS alloys possessed good corrosion resistances in both conditions, and better tensile properties. The applicability of the ADSS alloys as ATF cladding materials was discussed in view of their corrosion resistance, irradiation resistance, and embrittlement.

Original languageEnglish
Pages (from-to)1-14
Number of pages14
JournalJournal of Nuclear Materials
Publication statusPublished - 15 Aug 2018

Bibliographical note

Funding Information:
This study was mainly supported by the KUSTAR-KAIST International joint research project ( N11180043 ), and the MS&ICT/NRF ERC project ( 2016R1A5A1013919 ) of the Republic of Korea. The first author was supported by the Global Ph.D. Fellowship Program of the MS&ICT/NRF of the Republic of Korea ( 2015H1A2A1029835 ). The authors would also like to express their sincere appreciation to Lee Sung Yong and Jea Young Lim of the KEPCO Nuclear Fuel Co., for their help with the steam TG and PWR autoclave loop experiments.

Publisher Copyright:
© 2018 Elsevier B.V.


  • Accident-tolerant fuel cladding
  • Alumina-forming duplex stainless steel
  • High-temperature steam oxidation
  • PWR autoclave corrosion
  • Tensile strength

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Materials Science(all)
  • Nuclear Energy and Engineering


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