Research progress in high-temperature thermo-mechanical behavior for modelling FeCrAl cladding under loss-of-coolant accident condition

被引:5
作者
Qian, Libo [1 ]
Liu, Yu [1 ]
Huang, Tao [1 ]
Chen, Wei [1 ]
Du, Sijia [1 ]
Yin, Chunyu [1 ]
Xiong, Qingwen [1 ]
机构
[1] Nucl Power Inst China, Sci & Technol Reactor Syst Design Technol Lab, Chengdu 610213, Sichuan, Peoples R China
关键词
Review and reevaluation; FeCrAl cladding; High-temperature mechanical behavior; CANDIDATE MATERIALS; OXIDATION BEHAVIOR; NORMAL OPERATION; TOLERANT FUELS; ALLOYS; CREEP; STEAM; STEEL; AIR; DEFORMATION;
D O I
10.1016/j.pnucene.2023.104848
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
FeCrAl cladding has been proposed as one of the most promising accident tolerant fuel cladding candidates since the Fukushima Daiichi accident due to its excellent high-temperature mechanical properties, outstanding high-temperature steam oxidation resistance and low radiation-induced swelling rate. Cladding behavior under loss-of-coolant accident (LOCA) conditions would significantly affect reactor safety. Hence, the main focus of the present study is to provide fundamental safety analysis models for FeCrAl cladding under LOCA conditions. First, the fundamental models involved in LOCA safety analysis are identified, including high-temperature oxidation model, high-temperature failure model, high-temperature creep model, and high-temperature burst model. Second, by evaluating the research progress of high-temperature oxidation studies, the Robb model is recommended to estimate the oxidation rate of FeCrAl cladding, and the model of maximum tolerable temperature is also proposed. Then, the high-temperature creep and burst models of FeCrAl cladding are subsequently developed respectively by summarizing and reevaluating of high-temperature creep and burst data. Additionally, it has been verified that FeCrAl cladding has a much smaller burst strain than that of Zircaloy cladding, which may result in much smaller flow blockage, is verified by analyzing the deformation of FeCrAl cladding in burst experiment.
引用
收藏
页数:15
相关论文
共 92 条
[71]   Oxidation growth stresses in an alumina-forming ferritic steel measured by creep deflection [J].
Saunders, SRJ ;
Evans, HE ;
Li, M ;
Gohil, DD ;
Osgerby, S .
OXIDATION OF METALS, 1997, 48 (3-4) :189-200
[72]   Processability evaluation of a Mo-containing FeCrAl alloy for seamless thin wall tube fabrication [J].
Sun, Zhiciian ;
Yamamoto, Yukinori .
MATERIALS SCIENCE AND ENGINEERING A-STRUCTURAL MATERIALS PROPERTIES MICROSTRUCTURE AND PROCESSING, 2017, 700 :554-561
[73]  
Sweet R.T., 2018, Nuclear Engineering
[74]   Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam [J].
Tang, Chongchong ;
Jianu, Adrian ;
Steinbrueck, Martin ;
Grosse, Mirco ;
Weisenburger, Alfons ;
Seifert, Hans Juergen .
JOURNAL OF NUCLEAR MATERIALS, 2018, 511 :496-507
[75]   Advanced oxidation-resistant iron-based alloys for LWR fuel cladding [J].
Terrani, K. A. ;
Zinkle, S. J. ;
Snead, L. L. .
JOURNAL OF NUCLEAR MATERIALS, 2014, 448 (1-3) :420-435
[76]  
Terrani K A., 2016, Input Correlations for Irradiation Creep of FeCrAl and SiC Based on In-Pile Halden Test Results
[77]   Accident tolerant fuel cladding development: Promise, status, and challenges [J].
Terrani, Kurt A. .
JOURNAL OF NUCLEAR MATERIALS, 2018, 501 :13-30
[78]  
Tong L.S., 1986, The developing trends and thermal-hydraulic design of light water reactors
[79]   Alloy design and characterization of a recrystallized FeCrAl-ODS cladding for accident-tolerant BWR fuels: An overview of research activity in Japan [J].
Ukai, S. ;
Sakamoto, K. ;
Ohtsuka, S. ;
Yamashita, S. ;
Kimura, A. .
JOURNAL OF NUCLEAR MATERIALS, 2023, 583
[80]   High-temperature creep deformation in FeCrAl-oxide dispersion strengthened alloy cladding [J].
Ukai, S. ;
Kato, S. ;
Furukawa, T. ;
Ohtsuka, S. .
MATERIALS SCIENCE AND ENGINEERING A-STRUCTURAL MATERIALS PROPERTIES MICROSTRUCTURE AND PROCESSING, 2020, 794