Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition

Guo Zixuan; Jian Xiaobin; Li Wenjie; Zhang Kun; Wang Peng; Wang Yanpei

doi:10.13832/j.jnpe.2021.06.0254

Volume 42 Issue 6

Dec. 2021

Turn off MathJax

Article Contents

Article Navigation > Nuclear Power Engineering > 2021 > 42(6): 254-260

Guo Zixuan, Jian Xiaobin, Li Wenjie, Zhang Kun, Wang Peng, Wang Yanpei. Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition[J]. Nuclear Power Engineering, 2021, 42(6): 254-260. doi: 10.13832/j.jnpe.2021.06.0254

Citation:

Guo Zixuan, Jian Xiaobin, Li Wenjie, Zhang Kun, Wang Peng, Wang Yanpei. Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition[J]. Nuclear Power Engineering, 2021, 42(6): 254-260. doi: 10.13832/j.jnpe.2021.06.0254

Citation:

Guo Zixuan, Jian Xiaobin, Li Wenjie, Zhang Kun, Wang Peng, Wang Yanpei. Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition[J]. Nuclear Power Engineering, 2021, 42(6): 254-260. doi: 10.13832/j.jnpe.2021.06.0254

PDF( 3240 KB)

Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition

doi: 10.13832/j.jnpe.2021.06.0254

1.
Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China
2.
Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai, 200433, China

Received Date: 2021-07-14
Rev Recd Date: 2021-08-26
Publish Date: 2021-12-09

Abstract

Abstract

In the paper, the models of irradiation performance and thermo-mechanical constitutive relations of U-10Mo/Zr monolithic fuel element were established. With the finite element method, numerical simulation of fuel element thermo-mechanical performance under steady heterogeneous irradiation condition was conducted, the distribution and evolution characteristics of temperature, strain and stress in U-10Mo/Zr monolithic fuel element were acquired and analyzed. The results showed that the thickness increment of the fuel pellet becomes largest near the interface between fuel pellet and cladding, predominantly affected by fuel irradiation creep. Under the low burn-up conditions, the simulation result of fuel pellet swelling at high temperature condition equals the irradiation test results at low temperature. There are stress concentrations in the corner area of the fuel pellet and the outer area of the cladding end surface.
- U-Mo fuel,
- Monolithic fuel element,
- Thermo-Mechanical performance,
- Research and test reactors,
- Finite element method.

FullText(HTML)

References(20)

References

[1]	CRAWFORD D C, HAYES S L, POWERS J J. VTR startup fuel paper for NFSM: INL/EXT-18-44673-Rev000[R]. Idaho: Idaho National Laboratory, 2018.
[2]	DRAGUNOV Y G, TRETIYAKOV I T, LOPATKIN A V, et al. MBIR multipurpose fast reactor – Innovative tool for the development of nuclear power technologies[J]. Atomic Energy, 2012, 113(1): 24-28. doi: 10.1007/s10512-012-9590-x
[3]	KIM Y S, HOFMAN G L, CHEON J S, et al. Fission induced swelling and creep of U-Mo alloy fuel[J]. Journal of Nuclear Materials, 2013, 437(1-3): 37-46. doi: 10.1016/j.jnucmat.2013.01.346
[4]	SHOUDY A A, MCHUGH W E, SILLIMAN M A. The effect of irradiation temperature and fission rate on the radiation stability of uranium-10 wt% molybdenum alloy[C]//Proceedings of Radiation Damage in Reactor Materials. Vienna: IAEA, 1963.
[5]	LÓPEZ M, PICCHETTI B, TABOADA H. Influence of temperature and compressive stress on the UMo/Zry-4 interdiffusion layer[J]. Progress in Nuclear Energy, 2017(94): 101-105. doi: 10.1016/j.pnucene.2016.10.006
[6]	MILLER G K, BURKES D E, WACHS D M. Modeling thermal and stress behavior of the fuel-clad interface in monolithic fuel mini-plates[J]. Materials & Design, 2010, 31(7): 3234-3243.
[7]	YUN D, HOFMAN G L, KIM Y S, et al. Finite element modeling of irradiation induced swelling and creep in metallic mini-plate fuel - A preliminary study[J]. Transactions of the American Nuclear Society, 2011(105): 407-408.
[8]	ZHAO Y M, GONG X, DING S R. Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic U-Mo/Zr fuel plates under a heterogeneous irradiation condition[J]. Nuclear Engineering and Design, 2015(285): 85-97.
[9]	殷明阳,庞华,唐昌兵,等. UMo-Zr单片式燃料板结构改进研究[J]. 核动力工程,2019, 40(4): 172-176.
[10]	孔祥喆,丁淑蓉,田旭. UMo/Zr单片式燃料板堆内热力耦合行为研究[J]. 核动力工程,2018, 39(2): 109-113.
[11]	CUI Y, DING S R, CHEN Z T, et al. Modifications and applications of the mechanistic gaseous swelling model for UMo fuel[J]. Journal of Nuclear Materials, 2015(457): 157-164. doi: 10.1016/j.jnucmat.2014.11.065
[12]	WILLARD R M, SCHMITT A R. Irradiation swelling, phase reversion, and intergranular cracking of U-10wt. % Mo fuel alloy: NAA-SR-8956[R]. California: Atomics International, 1964.
[13]	BLEIBERG M L. Effect of fission rate and lamella spacing upon the irradiation-induced phase transformation of U-9wt% Mo alloy[J]. Journal of Nuclear Materials, 1959, 1(2): 182-190. doi: 10.1016/0022-3115(59)90051-0
[14]	KIM Y S, HOFMAN G L. Fission product induced swelling of U-Mo alloy fuel[J]. Journal of Nuclear Materials, 2011, 419(1-3): 291-301. doi: 10.1016/j.jnucmat.2011.08.018
[15]	YAN F, JIAN X B, DING S R. Effects of UMo irradiation creep on the thermo-mechanical behavior in monolithic UMo/Al fuel plates[J]. Journal of Nuclear Materials, 2019(524): 209-217. doi: 10.1016/j.jnucmat.2019.07.006
[16]	FISHER F E, RENKEN J C. Single-crystal elastic moduli and the Hcp→bcc transformation in Ti, Zr, and Hf[J]. Physical Review A, 1964, 135(2A): A482-A494. doi: 10.1103/PhysRev.135.A482
[17]	HAGRMAN D L, REYMAN G A. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior: NUREG/CR-0497[R]. Idaho: Idaho National Engineering Laboratory, 1979.
[18]	HALES J D, WILLIAMSON R L, NOVASCONE S R, et al. BISON theory manual the equations behind nuclear fuel analysis: INL/EXT-13-29930[R]. Idaho: Idaho National Laboratory, 2016.
[19]	JAEGER W. Heat transfer to liquid metals with empirical models for turbulent forced convection in various geometries[J]. Nuclear Engineering and Design, 2017(319): 12-27. doi: 10.1016/j.nucengdes.2017.04.028
[20]	REST J, KIM Y S, HOFMAN G L, et al. U-Mo Fuels Handbook: ANL-09/31[R]. Argonne: Argonne National Laboratory, 2006.