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Research Papers: Design and Analysis

Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure

[+] Author and Article Information
Jianfeng Mao

Engineering Research Center of Process Equipment and Re-manufacturing,
Ministry of Education,
Institute of Process Equipment and
Control Engineering,
Zhejiang University of Technology,
Chaowang Road 18#,
Hangzhou 310032, Zhejiang, China
e-mail: maojianfeng@zjut.edu.cn

Jianwei Zhu

Department of Mechanical and
Electrical Engineering,
Institute of Process Equipment and
Control Engineering,
Zhejiang University of Technology,
Huzhou Vocational & Technical College,
Chaowang Road 18#,
Hangzhou 310032, Zhejiang, China
e-mail: stormflash1978@163.com

Shiyi Bao

Institute of Process Equipment and
Control Engineering,
Zhejiang University of Technology,
Chaowang Road 18#,
Hangzhou 310032, Zhejiang, China
e-mail: bsy@zjut.edu.cn

Lijia Luo

Institute of Process Equipment and
Control Engineering,
Zhejiang University of Technology,
Chaowang Road 18#,
Hangzhou 310032, Zhejiang, China
e-mail: lijialuo@zjut.edu.cn

Zengliang Gao

Engineering Research Center of Process Equipment and Re-manufacturing,
Ministry of Education,
Institute of Process Equipment and
Control Engineering,
Zhejiang University of Technology,
Chaowang Road 18#,
Hangzhou 310032, Zhejiang, China
e-mail: zlgao@zjut.edu.cn

1Corresponding authors.

Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received February 17, 2016; final manuscript received August 22, 2016; published online September 28, 2016. Assoc. Editor: Reza Adibiasl.

J. Pressure Vessel Technol 139(2), 021206 (Sep 28, 2016) (8 pages) Paper No: PVT-16-1023; doi: 10.1115/1.4034582 History: Received February 17, 2016; Revised August 22, 2016

The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

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References

Figures

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Fig. 2

FE-model of the RPV with applied boundary condition

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Fig. 3

Inner temperature and wall thickness profiles used in FEM under critical heat flux

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Fig. 4

The distribution of external Mises stress along the thinnest wall

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Fig. 5

Maps of equivalent plastic strain distribution before creep occurrence

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Fig. 6

Comparison of equivalent stress distributions along the Path 1 with various internal pressures

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Fig. 7

Comparison of equivalent stress distributions along the Path 2 with various internal pressures

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Fig. 9

Maps of maximum principal stress distribution on the outside after 100 h creep hours

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Fig. 10

Contours of equivalent creep strain distribution after 100 h creep time

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Fig. 11

Comparison of total displacement distribution between 0 h and 100 h creep time among highly eroded region

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Fig. 8

Maps of maximum principal stress distribution after 100 h creep hours

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Fig. 12

Comparison of total damage distribution along external surface for various internal pressures

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Fig. 13

Comparison of total damage distribution along inner surface for various internal pressures

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Fig. 14

Contour plots of total damage distribution for the region of interest after 100 h

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Fig. 1

The scheme of RPV in core meltdown scenario and its cut vessel segment

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