船用堆燃料棒包壳疲劳寿命分析

李飞; 彭蕾; 时靖谊; 马冰; 金成; 解尧

doi:10.11889/j.0253-3219.2017.hjs.40.050606

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船用堆燃料棒包壳疲劳寿命分析

核能科学与工程 | 更新时间：2021-05-19

- 船用堆燃料棒包壳疲劳寿命分析
- Fatigue lifetime analysis on fuel cladding in marine reactor
- 核技术 2017年40卷第5期文章编号：050606
- 作者机构：
  
  中国科学技术大学核科学技术学院合肥 230027
- 作者简介：
  
  [ "李飞, 男, 1986年出生, 2008年毕业于海军工程大学, 现为硕士研究生, 从事裂变堆结构材料服役寿命分析", "LI Fei, male, born in 1986, graduated from Naval University of Engineering in 2008, master student, focusing on life analysis of fission reactor structure materials" ]
  彭蕾, E-mail:penglei@ustc.edu.cn PENG Lei, E-mail:penglei@ustc.edu.cn
- 基金信息：
  
  国家自然科学基金(No.11375173)
- DOI：10.11889/j.0253-3219.2017.hjs.40.050606
  中图分类号： TL341
- 收稿日期：2017-02-20，
  
  修回日期：2017-03-09，
  
  纸质出版日期：2017-05-10
- 稿件说明：
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李飞, 彭蕾, 时靖谊, 等. 船用堆燃料棒包壳疲劳寿命分析[J]. 核技术, 2017,40(5):050606

Fei LI, Lei PENG, Jingyi SHI, et al. Fatigue lifetime analysis on fuel cladding in marine reactor[J]. Nuclear techniques, 2017, 40(5): 050606
李飞, 彭蕾, 时靖谊, 等. 船用堆燃料棒包壳疲劳寿命分析[J]. 核技术, 2017,40(5):050606 DOI： 10.11889/j.0253-3219.2017.hjs.40.050606.

Fei LI, Lei PENG, Jingyi SHI, et al. Fatigue lifetime analysis on fuel cladding in marine reactor[J]. Nuclear techniques, 2017, 40(5): 050606 DOI： 10.11889/j.0253-3219.2017.hjs.40.050606.

摘要

船用堆瞬态变工况下燃料棒包壳温度和冷却剂压力波动较大，引起包壳的疲劳损伤，因此包壳疲劳寿命分析至关重要。本文利用ANSYS软件模拟船用堆瞬态变工况下燃料棒的热机械行为，结合锆包壳疲劳寿命设计曲线，考察包壳温度、冷却剂压力、燃料棒内压以及辐照对船用堆燃料棒包壳疲劳寿命的影响。计算结果表明，瞬态变工况使得包壳疲劳寿命有很大降低；包壳温度变化与冷却剂压力变化相比，前者对包壳疲劳寿命的影响小；辐照会降低包壳疲劳寿命。在不影响核动力船舶机动性的前提下，可采取一些必要的措施来降低包壳的疲劳损伤。

Abstract

Background

The maneuverability of nuclear-powered ships requires the sharp step change of the output power of the marine reactor

which causes fluctuation of fuel cladding temperature and coolant pressure. Since the enormous fluctuation leads to fatigue damage of cladding

it is important to analyze the fatigue lifetime of the fuel claddings in marine reactor. Purpose:The purpose is to analyze the effects of cladding temperature

coolant pressure

internal pressure of fuel rod and irradiation on the fatigue lifetime.

Methods

By using ANSYS

the thermalmechanical behaviors of fuel rods in transient operation condition were simulated. The fatigue lifetime of cladding under different conditions was investigated

based on the design curves of the fatigue lifetime for zircaloy cladding.

Results

The results indicate that the fatigue lifetime of the fuel cladding in marine reactor will reduce much in transient operation condition and irradiation can reduce fatigue lifetime of fuel cladding. The change of cladding temperature has little effect on the fatigue lifetime of cladding

compared with the change of coolant pressure.

Conclusion

The fatigue damage in the fuel cladding in marine reactor is very serious. Suitable measures are suggested to take in order to reduce the fatigue damage of the cladding

under the premise of ensuring the maneuverability of ships.

关键词

Keywords

references

蔡章生 . 核反应堆热工与安全分析 , : 武汉海军工程大学 , 2005 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Zhangsheng CAI . Thermal and safety analysis of nuclear reactor , : Wuhan Naval University of Engineering , 2005 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

彭敏俊 . 船舶核动力装置 , : 北京原子能出版社 , 2009 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Minjun PENG . Ship nuclear power plant , : Beijing Atomic Energy Press , 2009 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

王丹 , 王金雨 . 24个月换料周期燃料管理初步研究 . 核动力工程 , 2013 . 34 ( 增1 ): 99 - 102 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG2013S1023.htm .

Dan WANG , Jinyu WANG . Preliminary study on 24-month fuel management strategy . Nuclear Power Engineering , 2013 . 34 ( Suppl 1 ): 99 - 102 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG2013S1023.htm .

张炎 . 核动力舰船 . 国外核新闻 , 2005 . ( 4 ): 6 - 8 . http://www.cnki.com.cn/Article/CJFDTOTAL-SCHJ201505019.htm .

Yan ZHANG . Nuclear-powered ships . Foreign Nuclear News , 2005 . ( 4 ): 6 - 8 . http://www.cnki.com.cn/Article/CJFDTOTAL-SCHJ201505019.htm .

孙伟明 , 沈妙其 , 任欣 , . Zr-4合金常温疲劳性能的试验研究 . 核动力工程 , 1999 . 20 ( 3 ): 248 - 251 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG903.012.htm .

Weiming SUN , Miaoqi SHEN , Xin REN , . Test study on fatigue behavior of Zr-4 in room temperature condition . Nuclear Power Engineering , 1999 . 20 ( 3 ): 248 - 251 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG903.012.htm .

贾琦 , 蔡力勋 , 包陈 , . 锆合金薄壁细管的单轴拉伸与低周疲劳试验研究 . 原子能科学技术 , 2010 . 44 ( 6 ): 712 - 717 . http://www.cnki.com.cn/Article/CJFDTOTAL-YZJS201006014.htm .

Qi JIA , Lixun CAI , Chen BAO , . Uniaxial tension and low cycle fatigue behavior of thin-walled tubes for Zr-1Nb and Zr-4 . Atomic Energy Science and Technology , 2010 . 44 ( 6 ): 712 - 717 . http://www.cnki.com.cn/Article/CJFDTOTAL-YZJS201006014.htm .

陈彭 . 秦山核电厂燃料棒PCMI效应和包壳疲劳分析 . 原子能科学技术 , 1998 . 32 ( 2 ): 115 - 121 . http://www.cnki.com.cn/Article/CJFDTOTAL-YZJS802.003.htm .

Peng CHEN . Analysis on pellet-cladding mechanical interaction and cladding fatigue of the fuel rod of Qinshan nuclear power plant . Atomic Energy Science and Technology , 1998 . 32 ( 2 ): 115 - 121 . http://www.cnki.com.cn/Article/CJFDTOTAL-YZJS802.003.htm .

刘家正 , 张裕林 . 秦山一期核电站燃料组件加深燃耗后的燃料棒包壳应力、应变分析 . 核电工程与技术 , 2005 . 18 ( 3 ): 11 - 18 . http://cdmd.cnki.com.cn/Article/CDMD-10611-2009046710.htm .

Jiazheng LIU , Yulin ZHANG . Analysis of the stress and strain in the cladding for Qinshan PWR fuel assembly in extended burnup . Nuclear Power Engineering and Technology , 2005 . 18 ( 3 ): 11 - 18 . http://cdmd.cnki.com.cn/Article/CDMD-10611-2009046710.htm .

吕华权 , 刘彤 , 焦拥军 , . AFA 3G与AFA 2G燃料棒性能分析 . 核动力工程 , 2002 . 23 ( 5 ): 58 - 61 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG200205014.htm .

Huaquan LYU , Tong LIU , Yongjun JIAO , . AFA 2G and AFA 3G fuel rod performance analysis . Nuclear Power Engineering , 2002 . 23 ( 5 ): 58 - 61 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG200205014.htm .

Y J Jiang , Y Cui , Y Z Huo , . Three-dimensional FE analysis of the thermal-mechanical behaviors in the nuclear fuel rods . Annals of Nuclear Energy , 2011 . 38 ( 11 ): 2581 - 2593 . DOI: 10.1016/j.anucene.2011.07.016 http://doi.org/10.1016/j.anucene.2011.07.016 .

Yamaji A, Shimazaki J, Ochiai M, et al. The MRX integral reactor:maintenance and cost evaluation for ship application[R]. IAEA-TECDOC-977, Vienna, Austria:International Atomic Energy Agency, 1997.

T Kusunoki , N Odano , T Yoritsune , . Design of advanced integral-type marine reactor, MRX . Nuclear Engineering and Design , 2000 . 201 ( 2 ): 155 - 175 . http://www.ingentaconnect.com/content/els/00295493/2000/00000201/00000002/art00285 .

Z Vizvary , P Karditsas . Thermal fatigue of DEMO first wall due to pulsed operation . Fusion Engineering and Design , 2011 . 86 ( 9 ): 1587 - 1590 . https://www.researchgate.net/publication/251604508_Thermal_fatigue_of_DEMO_first_wall_due_to_pulsed_operation .

孙汉虹 , 程平东 , 缪鸿兴 , . 第三代核电技术AP1000 , : 北京中国电力出版社 , 2016 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Hanhong SUN , Pingdong CHENG , Hongxing MIAO , . The third generation nuclear power technology:AP1000 , : Beijing China Electric Power Press , 2016 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

B M Ma . Transient pellet-cladding interaction of LWR fuel rod-computer code ISUNE-4 . Nuclear Engineering and Design , 1980 . 58 ( 3 ): 303 - 338 . DOI: 10.1016/0029-5493(80)90146-6 http://doi.org/10.1016/0029-5493(80)90146-6 .

陈文振 , 于雷 , 郝建立 . 核动力装置热工水力 , : 北京中国原子能出版社 , 2013 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Wenzhen CHEN , Lei YU , Jianli HAO . Thermal hydraulics of nuclear power plant , : Beijing China Atomic Energy Press , 2013 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Cognet G, Efanov A, Fortov V, et al. Thermophysical properties database of materials for light water reactors and heavy water reactors[R]. IAEA-TECDOC-1496, Vienna, Austria:International Atomic Energy Agency, 2006.

A Soba , A Denis . Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods . Journal of Nuclear Materials , 2008 . 374 ( 1 ): 32 - 43 . http://www.academia.edu/7907284/Simulation_with_DIONISIO_1.0_of_thermal_and_mechanical_pellet-cladding_interaction_in_nuclear_fuel_rods .

J C Ramirez , M Stan , P Cristea . Simulations of heat and oxygen diffusion in UO 2 nuclear fuel rods . Journal of Nuclear Materials , 2006 . 359 ( 3 ): 174 - 184 . DOI: 10.1016/j.jnucmat.2006.08.018 http://doi.org/10.1016/j.jnucmat.2006.08.018 .

W J O'Donnell , B F Langer . Fatigue design basis for zircaloy components . Nuclear Science and Engineering , 1964 . 20 ( 1 ): 1 - 12 .

Bratton R N, Jessee M A, Wieselquist W A, et al. Rod internal pressure quantification and distribution analysis using FRAPCON[R]. Oak Ridge, TN, United States:Oak Ridge National Laboratory (ORNL), 2015.

李聪 , 应诗浩 , 沈保罗 . 水冷动力堆用锆合金的疲劳 . 核动力工程 , 2003 . 24 ( 1 ): 47 - 50 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG200301011.htm .

Cong LI , Shihao YING , Baoluo SHEN . Fatigue of zirconium alloys used in light water reactor . Nuclear Power Engineering , 2003 . 24 ( 1 ): 47 - 50 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG200301011.htm .

M K Samal , B K Dutta , S Guin , . A finite element program for on-line life assessment of critical plant components . Engineering Failure Analysis , 2009 . 16 ( 1 ): 85 - 111 . DOI: 10.1016/j.engfailanal.2008.01.007 http://doi.org/10.1016/j.engfailanal.2008.01.007 .

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