• Title/Summary/Keyword: Reactor Head

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1-D Two-phase Flow Investigation for External Reactor Vessel Cooling (원자로 용기 외벽냉각을 위한 1차원 이상유동 실험 및 해석)

  • Kim, Jae-Cheol;Park, Rae-Joon;Cho, Young-Rho;Kim, Sang-Baik;Kim, Sin;Ha, Kwang-Soon
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.31 no.5
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    • pp.482-490
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    • 2007
  • When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ${\pm}$15% error bounds and depended on the form loss due to the inlet/outlet area.

Study on bidirectional fluid-solid coupling characteristics of reactor coolant pump under steady-state condition

  • Wang, Xiuli;Lu, Yonggang;Zhu, Rongsheng;Fu, Qiang;Yu, Haoqian;Chen, Yiming
    • Nuclear Engineering and Technology
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    • v.51 no.7
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    • pp.1842-1852
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    • 2019
  • The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

Numerical study on thermal-hydraulics of external reactor vessel cooling in high-power reactor using MARS-KS1.5 code: CFD-aided estimation of natural circulation flow rate

  • Song, Min Seop;Park, Il Woong;Kim, Eung Soo;Lee, Yeon-Gun
    • Nuclear Engineering and Technology
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    • v.54 no.1
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    • pp.72-83
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    • 2022
  • This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

Computational Performance Prediction of Main Coolant Pump for the Integral Reactor SMART (일체형원자로 SMART 냉각재 순환펌프의 전산성능예측)

  • Kim M. H;Lee J. S;Park J. S;Kim J. I;Kim K. K
    • Journal of computational fluids engineering
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    • v.8 no.3
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    • pp.32-40
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    • 2003
  • CFD analyses of the three-dimensional turbulent flow in the impeller and diffuser of an axial flow pump including suction and discharge parts are presented and compared with experimental data. The purpose of the current study is to validate the CFD method for the performance analysis of the main coolant pump for SMART and to investigate the effect of suction and discharge shapes on the pump performance. To generate a performance curve, not only the design point but also the off-design points were computed. The results were compared with available experimental data in terms of head generated. At the design point, the analysis accurately predicts the experimental head value. In the range of the higher flow rates, the results are also in very good agreement with the experimental data, in magnitude but also in terms of slope of variation. For lower flow rates, the results shows that the analysis considering the suction and discharge well describe the typical S-shape performance curve of the axial pump.

Ultrasonic Phased Array Techniques for Detection of Flaws of Stud Bolts in Nuclear Power Plants

  • Lee, Joon-Hyun;Choi, Sang-Woo
    • Journal of the Korean Society for Nondestructive Testing
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    • v.26 no.6
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    • pp.440-446
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    • 2006
  • The reactor vessel body and closure head are fastened with the stud bolt that is one of crucial parts for safety of the reactor vessels in nuclear power plants. It is reported that the stud bolt is often experienced by fatigue cracks initiated at threads. Stud bolts are inspected by the ultrasonic technique during the overhaul periodically for the prevention of failure which leads to radioactive leakage from the nuclear reactor. The conventional ultrasonic inspection for stud bolts was mainly conducted by reflected echo method based on shadow effect. However, in this technique, there were numerous spurious signals reflected from every oblique surfaces of the thread. In this study, ultrasonic phased array technique was applied to investigate detectability of flaws in stud bolts and characteristics of ultrasonic images corresponding to different scanning methods, that is, sector and linear scan. For this purpose, simplified stud bolt specimens with artificial defects of various depths were prepared.

Qualification Test of a Main Coolant Pump for SMART Pilot (SMART 연구로 주냉각재펌프의 검증시험)

  • Park, Sang-Jin;Yoon, Eui-Soo;Oh, Hyoung-Woo
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.30 no.9 s.252
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    • pp.858-865
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    • 2006
  • SMART Pilot is a multipurpose small capacity integral type reactor. Main coolant pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of $310^{\circ}C$ and 14.7MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present wort a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and lift-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP.