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CFD Simulation of thermal hydraulic characteristics in a typical upper plenum of RPV

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Abstract

A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the Salome platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

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Abbreviations

CFD:

Computational fluid dynamics

PWR:

Pressurized water reactor

SC:

Support column

CRGT:

Control rod guide tube

FA:

Fuel assembly

RPV:

Reactor pressurized vessel

EPR:

European pressurized water reactor

VIPRE:

Versatile internals and component program for reactors, EPRI

ROCOM:

Rossendorf coolant mixing

NPP:

Nuclear power plant

RCCA:

Reactor core control assembly

COBRA:

Coolant boiling in rod arrays

SSG:

Speziale, Sarkar and Gatski

SST:

Shear-stress transport

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Acknowledgements

This work was supported by the National Natural Science Foundation of China (Grant No. 12075185).

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Authors and Affiliations

  1. State Key Laboratory of Multiphase Flow in Power Engineering, Department of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049, China

    Mingjun Wang, Lianfa Wang, Yingjie Wang, Wenxi Tian, Guanghui Su & Suizheng Qiu

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

    Jian Deng

Authors
  1. Mingjun Wang
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  2. Lianfa Wang
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  3. Yingjie Wang
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  4. Wenxi Tian
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  5. Jian Deng
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  6. Guanghui Su
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  7. Suizheng Qiu
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Correspondence to Wenxi Tian or Jian Deng.

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Wang, M., Wang, L., Wang, Y. et al. CFD Simulation of thermal hydraulic characteristics in a typical upper plenum of RPV. Front. Energy 15, 930–945 (2021). https://doi.org/10.1007/s11708-021-0728-1

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  • DOI: https://doi.org/10.1007/s11708-021-0728-1

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