Research on High Temperature Oxidation Behavior of Zirconium Alloy for Fuel Element Based on MOOSE Platform

Wu Zhouzhi; Zhang Kun; Wang Yanpei; Yu Hongxing; Zhang Lin; He Liang; Tang Changbing

doi:10.13832/j.jnpe.2024.01.0084

Volume 45 Issue 1

Feb. 2024

Turn off MathJax

Article Contents

Article Navigation > Nuclear Power Engineering > 2024 > 45(1): 84-89

Wu Zhouzhi, Zhang Kun, Wang Yanpei, Yu Hongxing, Zhang Lin, He Liang, Tang Changbing. Research on High Temperature Oxidation Behavior of Zirconium Alloy for Fuel Element Based on MOOSE Platform[J]. Nuclear Power Engineering, 2024, 45(1): 84-89. doi: 10.13832/j.jnpe.2024.01.0084

Citation:

Wu Zhouzhi, Zhang Kun, Wang Yanpei, Yu Hongxing, Zhang Lin, He Liang, Tang Changbing. Research on High Temperature Oxidation Behavior of Zirconium Alloy for Fuel Element Based on MOOSE Platform[J]. Nuclear Power Engineering, 2024, 45(1): 84-89. doi: 10.13832/j.jnpe.2024.01.0084

Citation:

Wu Zhouzhi, Zhang Kun, Wang Yanpei, Yu Hongxing, Zhang Lin, He Liang, Tang Changbing. Research on High Temperature Oxidation Behavior of Zirconium Alloy for Fuel Element Based on MOOSE Platform[J]. Nuclear Power Engineering, 2024, 45(1): 84-89. doi: 10.13832/j.jnpe.2024.01.0084

PDF( 8445 KB)

Research on High Temperature Oxidation Behavior of Zirconium Alloy for Fuel Element Based on MOOSE Platform

doi: 10.13832/j.jnpe.2024.01.0084

Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China

Received Date: 2022-12-29
Rev Recd Date: 2023-02-10
Publish Date: 2024-02-15

Abstract

Abstract

In order to establish a prediction method for the high-temperature oxidation behavior of the new N36 zirconium alloy and allow the autonomous fuel element performance analysis code FORWARD to be applied to the loss of coolant accident (LOCA) condition, the high-temperature steam oxidation test of the new N36 zirconium alloy was carried out in this study. The high-temperature oxidation model of N36 zirconium alloy was developed and validated, and the high-temperature oxidation behavior of N36 zirconium alloy under LOCA condition was predicted using the FORWARD code. The results show that the predicted oxidation weight gain of N36 zirconium alloy is in good agreement with the verification test results, and the predicted oxidation behavior of N36 zirconium alloy at high temperature under LOCA condition is reasonable. Therefore, the model and fuel element performance analysis code developed in this study can be used to predict the high temperature oxidation behavior of the new N36 zirconium alloy.
- N36 zirconium alloy,
- High temperature oxidation,
- MOOSE,
- FORWARD code

FullText(HTML)

References(14)

References

[1]	MEYER R O. An assessment of fuel damage in postulated reactivity-initiated accidents[J]. Nuclear Technology, 2006, 155(3): 293-311. doi: 10.13182/NT06-A3763
[2]	YOO H I, PARK S H. Experimental verification of a kinetic model of Zr-oxidation[J]. Journal of the Korean Ceramic Society, 2006, 43(11): 724-727. doi: 10.4191/KCERS.2006.43.11.724
[3]	邱军,赵文金,GUILBERT T,等. 3种锆合金的高温氧化行为[J]. 金属学报,2011, 47(9): 1216-1220.
[4]	LEISTIKOW S, SCHANZ G, ZUREK Z. Comparison of high temperature steam oxidation behavior of zircaloy-4 versus austenitic and ferritic steels under light water reactor safety aspects[C]//The Polish-German Seminar on Properties of High Temperature Alloys. Cracow, Poland: Materials Science, 1987
[5]	SCHANZ G. Recommendations and supporting information on the choice of zirconium oxidation models in severe accident codes: FZKA-6827[R]. Karlsruhe: Forschungszentrum Karlsruhe GmbH, 2003.
[6]	LEISTIKOW S, SCHANZ G, BERG H V, et al. Comprehensive presentation of extended Zircaloy-4 steam oxidation results 600-1600℃[C]//OECD-NEACSNI/IAEA Specialists Meeting on Water Reactor Fuel Safety and Fission Product Release in Off-Normal and Accident Conditions. Roskilde: Riso National Laboratory, 1983: 188-199.
[7]	PRATER J T, COURTRIGHT E L. Zircaloy-4 oxidation at 1300 to 2400℃: NUREG/CR-4889[R]. Richland: Pacific Northwest Laboratory, 1987: 489-501.
[8]	CATHCART J V, PAWEL R E, MCKEE R A, et al. Zirconium metal-water oxidation kinetics. IV. Reaction rate studies. [BWR: PWR]: ORNL/NUREG-17[R]. Oak Ridge: Oak Ridge National Lab, 1977.
[9]	GEELHOOD K J, LUSCHER W G, BEYER C E, et al. FRAPTRAN 1.4: a computer code for the transient analysis of oxide fuel rods[M]. Washington: US Nuclear Regulatory Commission, 2011: 1-6.
[10]	SINGH G, TERRANI K, KATOH Y. Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis[J]. Journal of Nuclear Materials, 2018, 499: 126-143. doi: 10.1016/j.jnucmat.2017.11.004
[11]	邓超群,向烽瑞,贺亚男,等. 基于MOOSE平台的棒状燃料元件性能瞬态分析程序开发与验证[J]. 原子能科学技术,2021, 55(8): 1429-1439. doi: 10.7538/yzk.2020.youxian.0607
[12]	SLAUGHTER A E, JOHNSON M J, TONKS M R, et al. MOOSE: a framework to enable rapid advances and collaboration in modeling snow and avalanches[C]//International Snow Science Workshop 2014 Proceedings. Banff: International Snow Science Workshop, 2014: 601-607.
[13]	GASTON D, NEWMAN C, HANSEN G, et al. MOOSE: a parallel computational framework for coupled systems of nonlinear equations[J]. Nuclear Engineering and Design, 2009, 239(10): 1768-1778. doi: 10.1016/j.nucengdes.2009.05.021
[14]	HALES D J, WILLIAMSON R L, NOVASCONE S R, et al. BISON theory manual the equations behind nuclear fuel analysis[R]. Idaho Falls: Idaho National Laboratory, 2016: 122-124.