Analysis of Thermal Flow Field for Reactor Core at Zero Power Steam Line Break Accident

2021 ◽  
Vol 9 (2) ◽  
pp. 97-106
Author(s):  
Chongkuk Chun
Keyword(s):  
Flow Field ◽  
Reactor Core ◽  
Steam Line ◽  
Thermal Flow

Numerical Investigation of Thermal Flow Field of Air-Cooled Condensers for Yongshou Plant

2012 ◽  
Vol 248 ◽  
pp. 391-394
Author(s):  
Wen Zhou Yan ◽  
Wan Li Zhao ◽  
Qiu Yan Li
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Wind Speed ◽  
Power Plant ◽  
Flow Field ◽  
Wind Direction ◽  
Rotational Speed ◽  
Power Plants ◽  
Thermal Flow ◽  
Air Cooled Condensers

By using the computational fluid dynamics code, FLUENT, Numerically simulation is investigated for Youngshou power plant. Under the constant ambient temperature, the effects of different wind speed and wind direction on the thermal flow field are qualitatively considered. It was found that when considering about the existing and normally operating power plants, the thermal flow field is more sensitive to wind direction and wind speed. Based on the above results, three improved measures such as: increasing the wind-wall height and accelerating the rotational speed of the fans near the edge of the ACC platform and lengthen or widen the platform are developed to effectively improving the thermal flow field, and enhanced the heat dispersal of ACC.

Analysis of Pressure Wave Phenomena in Steam Line Piping

2004 ◽  
Author(s):  
Juswald Vedovi ◽  
Michael J. Hibbard ◽  
Donald R. Todd ◽  
Kostadin Ivanov
Keyword(s):  
Pressure Wave ◽  
Reactor Core ◽  
Steam Line ◽  
Step Size ◽  
Time Step ◽  
Time Step Size ◽  
Bypass Line ◽  
Sensitivity Studies ◽  
Main Steam ◽  
Wave Phenomena

One challenging aspect of boiling water reactor (BWR) analysis is the ability to predict the system response following the rapid closure of a steam line control or isolation valve. Of particular interest is to accurately model the effect of a pressurization wave as it transits through the reactor core. This paper describes sensitivity studies, which were performed to demonstrate the predictive capabilities of S-RELAP5 for analysis of pressure wave phenomena, and it describes the steam line models developed in support of this effort. S-RELAP5 is a RELAP5-based thermal-hydraulic system code used for realistic analyses of large break loss-of-coolant accidents (LBLOCA) in pressurized water reactors (PWRs). The code is also suitable for analyzing PWR small break LOCA and non-LOCA transients. On the extent of the analyses documented in this paper, there are not significant differences between S-RELAP5 and RELAP5. Framatome ANP is developing code models which will extend the capability of S-RELAP5 to analyze BWR transients. Within the framework of this development work, a task was established to investigate the capability of S-RELAP5 to model transients involving steam line pressure wave phenomena. An additional goal of this task was development of a steam line nodalization guideline for modeling pressurization transients. To achieve these goals, various steam line models were investigated and a series of sensitivity studies were performed using the Peach Bottom Unit 2 Turbine Trip test series as an experimental benchmark. A Turbine Trip is an anticipated operational occurrence (AOO), which, for analysis purposes, is initiated by rapid closure of the Turbine Stop Valves (TSV). The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes to the core void distribution and coolant flow. The magnitude of the resulting neutron flux transient is significantly affected by the void collapse. The performed sensitivity studies focused on steam line geometry, time step size, steam line node length, Turbine Stop Valve (TSV) model, steam Bypass Valve (BPV), and transient boundary conditions. Three models were developed for the Peach Bottom Unit 2 steam line; a single ideal steam line with no bends or elevation changes, a four steam line model of the PB2 main steam and bypass line piping, and an equivalent single steam line model of the PB2 main steam and bypass line piping. The pressure response calculated by S-RELAP5 was compared to theoretical predictions based on fundamental water-hammer equations and to experimental data from the PB2 turbine trip tests (TT1, TT2, and TT3). The results demonstrate the capability of S-RELAP5 to accurately predict pressure wave phenomena. Additional results of this work include recommendations of appropriate values for time step size and steam line node size. Models for the TSV and BPV were developed based on TT2 information, and validated through analysis of TT1 and TT3. Finally, the results of this work demonstrate the strong interaction between the steam bypass line and the main steam line, and the corresponding impact on system pressure response.

Wind Effect, Recirculation and Thermal Flow Field of a Direct Air-cooled Condenser for a Large Power Plant

2011 ◽  
Author(s):  
W. L. Zhao ◽  
P. Q. Liu ◽  
H. S. Duan ◽  
J. Y. Zhu ◽  
Jiachun Li ◽  
...  
Keyword(s):  
Power Plant ◽  
Flow Field ◽  
Wind Effect ◽  
Thermal Flow ◽  
Large Power

Computational and Experimental Investigation of Thermal Flow Field of Micro Turbojet Engine with Various Nozzle Configurations

2018 ◽  
Vol 46 (2) ◽  
pp. 150-158 ◽  
Author(s):  
Hyun-Jin Lee ◽  
Ji-Hyun Lee ◽  
Rho-Shin Myong ◽  
Sun-Mi Kim ◽  
Sung-Man Choi ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Flow Field ◽  
Thermal Flow ◽  
Turbojet Engine

Simulation Study on the Influence of Rocket Tail Flame of Ejection Seat on the Thermal Flow Field in Engine Room

2021 ◽  
pp. 911-920
Author(s):  
Lou Jin ◽  
Yibin Wu ◽  
Liang Wu ◽  
Dongyue Cui ◽  
Pin Yang
Keyword(s):  
Flow Field ◽  
Simulation Study ◽  
Thermal Flow ◽  
Engine Room

VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

2016 ◽  
Author(s):  
Vefa N. Kucukboyaci ◽  
Yixing Sung ◽  
Yiban Xu ◽  
Liping Cao ◽  
Robert K. Salko
Keyword(s):  
Reactor Core ◽  
Steam Line ◽  
High Flow ◽  
Low Flow ◽  
Code System ◽  
Fuel Temperature ◽  
Transport Code ◽  
Main Steam Line ◽  
Main Steam ◽  
Flow Case

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics (T/H), and fuel temperature components with an isotopic depletion capability. The neutronics capability is based on the Michigan Parallel Characteristics Transport Code (MPACT), a three-dimensional whole-core transport code. The T/H and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to the departure from nucleate boiling (DNB) ratio at the most limiting point of a postulated pressurized water reactor main steam line break event initiated at the hot zero power, either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power, where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady-state reactor core response under the main steam line break accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

The Optimization of Thermal Flow Field in a Large-Size MOCVD Reactor

2013 ◽  
Vol 52 (1) ◽  
pp. 1021-1026 ◽  
Author(s):  
C.-K. Hu ◽  
T. T. Li ◽  
Y.-J. Lin
Keyword(s):  
Flow Field ◽  
Thermal Flow ◽  
Mocvd Reactor ◽  
Large Size
Sign in / Sign up

Export Citation Format

Share Document