Journal of Shanghai Jiao Tong University

Journal of Shanghai Jiaotong University >

2022 , Vol. 56 >Issue 1: 14 - 20

DOI: https://doi.org/10.16183/j.cnki.jsjtu.2020.395

Test of Critical Heat Flux on Outer Surface of CAP1400 Pressurized Reactor Vessel

Expand
  • 1. Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai 200233, China
    2. School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China

Received date: 2020-11-26

  Online published: 2022-01-21

Fold

Abstract

A full height sliced test facility is established based on the modeling process. The heating test section is manufactured by adopting the explosive welding technology, and the surface of carbon steel can simulate the actual surface characteristic of a pressurized reactor. Critical heat flux tests on flow, pressure, subcooling degree, chemical material in water, and surface characteristics are performed on the outer surface of the pressurized reactor vessel. Therefore, the law of heat transfer limit on the outer surface of the CAP1400 pressurized reactor vessel is obtained, and the process of flow and heat transfer for in-vessel retention is validated.

Key words： outer surface of pressurized reactor vessel; critical heat flux test; in-vessel retention

Cite this article

SHI Guobao, ZHENG Mingguang, ZHANG Kun, KUANG Bo, LIU Pengfei . Test of Critical Heat Flux on Outer Surface of CAP1400 Pressurized Reactor Vessel[J]. Journal of Shanghai Jiaotong University, 2022 , 56(1) : 14 -20 . DOI: 10.16183/j.cnki.jsjtu.2020.395

References

[1] 郑明光. 从AP1000到CAP1400, 我国先进三代非能动核电技术自主化历程[J]. 中国核电, 2018, 11(1):41-45.
[1] ZHENG Mingguang. From AP1000 to CAP1400, self-reliant process of China’s third generation nuclear power[J]. China Nuclear Power, 2018, 11(1):41-45.
[2] SHI G B, GU P W, LU W, et al. CAP1400 IVR related design features and assessment[J]. Nuclear Engineering and Design, 2019, 346:35-45.
[3] THEOFANOUS T G, LIU C, ADDITON S, et al. In-vessel coolability and retention of a core melt: TRN: 97: 012637 [R]. Office of Scientific and Technical Information (OSTI): U.S. Department of Energy Office of Scientific and Technical Information, 1996.
[4] THEOFANOUS T G, TU J P, SALMASSI T,, et al. Quantification of limits to coolability in ULPU-2000 configuration IV [R/OL]. (2002-06-06)[2020-06-15]. https://www.nrc.gov/docs/ML0216/ML021620559.pdf.
[5] DINH T N, TU J P, SALMASSI T, et al. Limits of the coolability in the AP1000-related ULPU-2400 configuration V facility[J]. International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), 2003: 1-14.
[6] LEE J, CHANG S H. An experimental study on CHF in pool boiling system with SA508 test heater under atmospheric pressure[J]. Nuclear Engineering and Design, 2012, 250:720-724.
[7] DEWITT G, MCKRELL T, BUONGIORNO J, et al. Experimental study of critical heat flux with alumina-water nanofluids in downward-facing channels for in-vessel retention applications[J]. Nuclear Engineering and Technology, 2013, 45(3):335-346.
[8] TROJER M, AZIZIAN R, PARAS J, et al. A margin missed: The effect of surface oxidation on CHF enhancement in IVR accident scenarios[J]. Nuclear Engineering and Design, 2018, 335:140-150.
[9] KAM D H, CHOI Y J, JEONG Y H. CHF experiment with downward-facing carbon and stainless steel plates under pressurized conditions[J]. International Journal of Heat and Mass Transfer, 2018, 125:670-680.
[10] SAKASHITA H, ONO A, NYUI J. Critical heat flux and near-wall boiling behaviors in saturated and subcooled pool boiling on vertical and inclined surfaces[J]. Journal of Nuclear Science and Technology, 2009, 46(11):1038-1048.
Options
Outlines

/

〈 〉