Isothermal experiments on steam oxidation of magnetron-sputtered chromium-coated zirconium alloy cladding at 1200 ℃

Degradation mechanism of SiCf/SiC composites after long-time water vapor and oxygen corrosion at 1300 °C

(2022) Jin Zhang et al.   CORROSION SCIENCE

Microstructure and tensile behavior of powder metallurgy FeCrAl accident tolerant fuel cladding

(2022) Shenyan Huang et al.   JOURNAL OF NUCLEAR MATERIALS

Development of oxidation model for zirconium alloy cladding and application in the analysis of cladding behavior under loss of coolant accident

(2022) Dong Wang et al.   JOURNAL OF NUCLEAR MATERIALS

Systematic investigations on the coating degradation mechanism during the steam oxidation of Cr-coated Zry-4 at 1200 °C

(2022) Junkai Liu et al.   CORROSION SCIENCE

High temperature Cr-Zr interaction of two types of Cr-coated Zr alloys in inert gas environment

(2021) Jianqiao Yang et al.   JOURNAL OF NUCLEAR MATERIALS

Primary radiation damage characteristics in displacement cascades of FeCrAl alloys

(2021) Tianzhou Ye et al.   JOURNAL OF NUCLEAR MATERIALS

Steam oxidation of Cr-coated Sn-containing Zircaloy solid rod at 1000 °C

(2021) Junkai Liu et al.   CORROSION SCIENCE

Development of low-Cr wrought FeCrAl cladding alloys and its irradiation tolerance and steam oxidation resistance at 1200 °C

(2021) Hui Wang et al.   CORROSION SCIENCE

Design and characterization of zirconium-based multilayer coating for carbon/carbon composites against oxyacetylene ablation

(2021) Guanghui Feng et al.   CORROSION SCIENCE

Exploring the high-temperature steam oxidation behaviors of the lean-Cr (7–10 wt%) FeCrAl alloys

(2021) Guang Chen et al.   CORROSION SCIENCE

Transient experiments on oxidation and degradation of Cr-coated Zircaloy in steam up to 1600 ℃

(2021) Junkai Liu et al.   CORROSION SCIENCE

High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage

(2021) M. Steinbrück et al.   JOURNAL OF NUCLEAR MATERIALS

Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan

(2021) K. Sakamoto et al.   JOURNAL OF NUCLEAR MATERIALS

“High Temperature Steam Oxidation of Chromium-Coated Zirconium-Based Alloys: Kinetics and Process”

(2020) Jean-Christophe Brachet et al.   CORROSION SCIENCE

Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility

(2020) J.C. Brachet et al.   JOURNAL OF NUCLEAR MATERIALS

A systematic study of the oxidation behavior of Cr coatings on Zry4 substrates in high temperature steam environment

(2020) Xiaochun Han et al.   CORROSION SCIENCE

Oxidation kinetics of Cr-coated zirconium alloy: Effect of coating thickness and microstructure

(2020) E.B. Kashkarov et al.   CORROSION SCIENCE

High-temperature oxidation of thick Cr coating prepared by arc deposition for accident tolerant fuel claddings

(2019) Xiaogang Hu et al.   JOURNAL OF NUCLEAR MATERIALS

Development of cold spray chromium coatings for improved accident tolerant zirconium-alloy cladding

(2019) Benjamin Maier et al.   JOURNAL OF NUCLEAR MATERIALS

An interesting oxidation phenomenon of Cr coatings on Zry-4 substrates in high temperature steam environment

(2019) Xiaochun Han et al.   CORROSION SCIENCE

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors

(2019) Jean-Christophe Brachet et al.   JOURNAL OF NUCLEAR MATERIALS

Microstructure, corrosion resistance and oxidation behavior of Cr-coatings on Zircaloy-4 prepared by vacuum arc plasma deposition

(2019) Tianguo Wei et al.   CORROSION SCIENCE

Development and testing of multicomponent fuel cladding with enhanced accidental performance

(2019) Jakub Krejčí et al.   Nuclear Engineering and Technology

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

(2019) Hwasung Yeom et al.   JOURNAL OF NUCLEAR MATERIALS

Numerical simulation of zircaloy-water reaction based on the moving particle semi-implicit method and combined analysis with the MIDAC code for the nuclear-reactor core melting process

(2019) Dong Wang et al.   PROGRESS IN NUCLEAR ENERGY

Accident tolerant fuel cladding development: Promise, status, and challenges

(2018) Kurt A. Terrani   JOURNAL OF NUCLEAR MATERIALS

Preliminary evaluation of U 3 Si 2 -FeCrAl fuel performance in light water reactors through a multi-physics coupled way

(2018) Yanan He et al.   NUCLEAR ENGINEERING AND DESIGN

A comparative study on preliminary performance evaluation of ATFs under normal and accident conditions with FRAP-ATF code

(2018) Bowen Qiu et al.   PROGRESS IN NUCLEAR ENERGY

Behavior of plasma sprayed Cr coatings and FeCrAl coatings on Zr fuel cladding under loss-of-coolant accident conditions

(2018) Yiding Wang et al.   SURFACE & COATINGS TECHNOLOGY

Cladding oxidation model development based on diffusion equations and a simulation of the monoclinic-tetragonal phase transformation of zirconia during transient oxidation

(2014) Xiaoqiang He et al.   JOURNAL OF NUCLEAR MATERIALS

Accident tolerant fuels for LWRs: A perspective

(2013) S.J. Zinkle et al.   JOURNAL OF NUCLEAR MATERIALS

Insights from review and analysis of the Fukushima Dai-ichi accident

(2012) Masashi Hirano et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

Oxidation kinetics and oxygen diffusion in low-tin Zircaloy-4 up to 1523K

(2008) X. Ma et al.   JOURNAL OF NUCLEAR MATERIALS