scholarly journals Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

2016 ◽  
Vol 48 (4) ◽  
pp. 941-951 ◽  
Author(s):  
Istvan Farkas ◽  
Ezddin Hutli ◽  
Tatiana Farkas ◽  
Antal Takács ◽  
Attila Guba ◽  
...  
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Pressurized Water Reactor ◽  
Reactor Model ◽  
Pressurized Water ◽  
Flow Measurements ◽  
Primary Loop ◽  
Water Reactor ◽  
Model Flow ◽  
Dynamics Calculation

scholarly journals SIMULAÇÃO DO ESCOAMENTO EM CONTRACORRENTE ÁGUA/AR EM REATORES NUCLEARES PWR UTILIZANDO CÓDIGO OPENFOAM

2018 ◽  
Vol 6 ◽  
pp. 92-102
Author(s):  
Youssef Morghi ◽  
Amir Zacarias Mesquita ◽  
Jesus Alfonso Puente Angulo ◽  
Ana Rosa Baliza Maia
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

A limitação do escoamento em contracorrente, ou inundação, é um fenômeno caracterizado pelo controle que um gás exerce no escoamento de um líquido em sentido contrário. Este efeito tem recebido atenção especial da área nuclear, devido à sua influência no comportamento termofluidodinâmico dos reatores nucleares refrigerados à água pressurizada (Pressurized Water Reactor – PWR), durante um acidente de perda de refrigerante – LOCA (Loss-of-Coolant Accident). A modelagem  numérica  constitui uma  ferramenta fundamental para o desenvolvimento da engenharia nuclear. Este trabalho tem o popósito de demonstrar que o software de dinâmica de fluídos computacional – CFD (Computational Fluid Dynamics) OpenFOAM®, tem potencial para ser utilizado na modelagem do fenômeno de escoamento em contracorrente que ocorre no núcleo dos reatores PWR. Uma introdução de CFD usando OpenFOAM. O solver utilizado foi o interFoam, onde foi observado que o openFOAM  apresentou resultados satisfatórios para o estudo de escoamento multifásicos.

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

2010 ◽  
Vol 240 (10) ◽  
pp. 2789-2799 ◽  
Author(s):  
B. Vogt ◽  
K. Fischer ◽  
J. Starflinger ◽  
E. Laurien ◽  
T. Schulenberg
Keyword(s):  
Supercritical Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Primary Loop ◽  
Water Reactor

Visual simulation of thermal fluid dynamics in a pressurized water reactor

2009 ◽  
Vol 25 (11) ◽  
pp. 985-996 ◽  
Author(s):  
Zhe Fan ◽  
Yu-Chuan Kuo ◽  
Ye Zhao ◽  
Feng Qiu ◽  
Arie Kaufman ◽  
...  
Keyword(s):  
Fluid Dynamics ◽  
Visual Simulation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Pressurized Water Reactor Control Rods Worth Calibration Calculation by MCNP

2014 ◽  
Vol 539 ◽  
pp. 684-687
Author(s):  
Bo Yang ◽  
He Xi Wu ◽  
Qiang Lin Wei ◽  
Yi Bao Liu
Keyword(s):  
Neutron Flux ◽  
Flux Density ◽  
Operating Time ◽  
Pressurized Water Reactor ◽  
Reactor Model ◽  
Density Peak ◽  
Pressurized Water ◽  
Water Reactor ◽  
Neutron Flux Density ◽  
Control Rods

Control rods play an important role in nuclear power plant's reactivity control. In this paper, the study first establishes the pressurized water reactor model with Control rods by MCNP program, calculates the reactor keff by KCODE card and neutron flux density by F5:N card. The result shows that when control rods are not inserted, the neutron flux density distribution is similar to the cosine function. The control rods slowly but continuously move up with the reactor's increasing operating time, the neutron flux density peak gradually shifted to the top of reactor core. The simulation results agree with the nuclear fuel management program.

Water hammer characteristics of integral pressurized water reactor primary loop

2013 ◽  
Vol 261 ◽  
pp. 165-173 ◽  
Author(s):  
Qiaolin Zuo ◽  
Suizheng Qiu ◽  
Wei Lu ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Water Hammer ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Primary Loop ◽  
Water Reactor

Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)

2017 ◽  
Author(s):  
Huasong Cao
Keyword(s):  
Steam Generator ◽  
System Response ◽  
Outlet Temperature ◽  
Pressurized Water Reactor ◽  
Reactor Model ◽  
Design Data ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Steam Generator Tube ◽  
Water Reactor

Lots of efforts have been made to Research & Development of Small pressurized water reactors (SPWRs). Steam generator tube break occurs due to wear and corrosion frequently in the reactor. Among the breaks, Small Steam Generator Tube Break (SSGTB) is difficult to detect. Therefore, it is necessary to investigate the features of SSGTB. A small pressurized water reactor model has been established in this paper by Relap5. The model includes reactor core, pressurizer, steam generator, main coolant pump and auxiliary safety system. The core flow, pressure of pressurizer, core outlet temperature and secondary outlet steam temperature obtained based on steady-state calculation is compared with design data to verify the model correct. SSGTB is simulated by introducing a small break in the steam generator tube. The important parameters of reactor are recorded and analyzed. The procedure of SSGTB is analyzed and the system response features are summarized.

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