Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 232-236 ◽  
Author(s):  
D. L. Zhang ◽  
P. Song ◽  
S. Wang ◽  
X. Wang ◽  
J. Chen ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Pool Type ◽  
Code Development

Study on the Heat Transfer in the Water Pool Type Reactor Cavity Cooling System of the Very High Temperature Gas Cooled Reactor

2005 ◽  
Author(s):  
H. K. Cho ◽  
D. U. Seo ◽  
M. O. Kim ◽  
G. C. Park
Keyword(s):  
Heat Transfer ◽  
Forced Convection ◽  
Cooling System ◽  
Reactor Vessel ◽  
Water Pool ◽  
Cooling Pipe ◽  
Cavity Cooling ◽  
The Heat Transfer Coefficient ◽  
Pool Type ◽  
High Temperature Gas

In the HTGR (High Temperature Gas Cooled Reactor), the Reactor Cavity Cooling System (RCCS) is equipped to remove the heat transferred from the reactor vessel to the structure of the containment. The function of the RCCS is to dissipate the heat from the reactor cavity during normal operation including shutdown. The system also removes the decay heat during the loss of forced convection (LOFC) accident. A new concept of the water pool type RCCS was proposed at Seoul National University. The system mainly consists of two parts, water pool located between the containment and reactor vessel and five trains of air cooling system installed in the water pool. In normal operations, the heat loss from the reactor vessel is transferred into the water pool via cavity and it is removed by the forced convection of air flowing through the cooling pipes. During the LOFC accident, the after heat is passively removed by the water tank without the forced convection of air and the RCCS water pool is designed to provide sufficient passive cooling capacity of the after heat removal for three days. In the present study, experiments and numerical calculations using CFX5.7 for the water pool and cooling pipe were performed to investigate the heat transfer characteristics and evaluate the heat transfer coefficient model of the MARS-GCR (Multi-dimensional Analysis of Reactor Safety for Gas Cooled Reactor Analysis) which was developed for the safety analysis of the gas cooled reactor. From the results of the experiments and CFX calculations, heat transfer coefficients inside the cooling pipe were calculated and those were used for the assessment for the heat transfer coefficient model of the MARS-GCR.

Preliminary Analysis of Unprotected Loss of Heat Sink in Small Long Life Gas Cooled Fast Reactor

2015 ◽  
Vol 751 ◽  
pp. 268-272
Author(s):  
Su'ud Zaki ◽  
Nuri Trianti ◽  
Rosidah M. Indah
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Channel Model ◽  
Cooling System ◽  
Hot Spot ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reducing Power ◽  
Secondary System ◽  
Heat Removal System

The failure of the secondary side in Gas Cooled Fast Reactor system, which may contain co-generation system, will cause loss of heat sink (LOHS) accident. In this study accident analysis of unprotected loss of heat sink due to the failure of the secondary cooling system has been investigated. The thermal hydraulic model include transient hot spot channel model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is also plays important role to limit the level of consequence during the accident. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate the failure of the secondary system by reducing power through reactivity feedback and remove the rest of heat through natural circulations based decay heat removal (RVACS system).

The design and experimental validation of an emergency core cooling system for a pool type research reactor

1999 ◽  
Vol 26 (8) ◽  
pp. 709-728 ◽  
Author(s):  
Walmir Maximo Torres ◽  
Benedito Dias Baptista Filho ◽  
Daniel Kao Sun Ting
Keyword(s):  
Experimental Validation ◽  
Research Reactor ◽  
Cooling System ◽  
Core Cooling ◽  
Pool Type

A CONCEPTUAL DESIGN STUDY OF POOL-TYPE SODIUM-COOLED FAST REACTOR WITH ENHANCED ANTI-SEISMIC CAPABILITY

2019 ◽  
Vol 2019.27 (0) ◽  
pp. 1651
Author(s):  
Shigenobu Kubo ◽  
Yoshitaka Chikazawa ◽  
Hiroyuki Ohshima ◽  
Masato Uchita ◽  
Takayuki Miyagawa ◽  
...  
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Design Study ◽  
Pool Type

Numerical Analysis of Flow-Induced Vibration of Large Diameter Pipe With Short Elbow

2017 ◽  
Author(s):  
Shigeru Takaya ◽  
Tatsuya Fujisaki ◽  
Masaaki Tanaka
Keyword(s):  
Numerical Simulation ◽  
Numerical Analysis ◽  
Fast Reactor ◽  
Pressure Fluctuation ◽  
Cooling System ◽  
Large Diameter ◽  
Recirculation Region ◽  
Flow Induced Vibration ◽  
Scaled Model ◽  
Energy Agency

Japan Atomic Energy Agency is now conducting design study and R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than a prototype fast reactor in Japan, Monju, in order to reduce construction costs and enhance economic performance. The design, however, increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design. Numerical simulation is a useful method to deal with such a complex phenomenon. We have been developing numerical analysis models of the hot-leg piping using Unsteady Reynolds Averaged Navier-Stokes simulation with Reynolds stress model. In this study, numerical simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment, but it was found some underestimation at other parts, especially in relatively high frequency range. Eigenvalue vibration analysis was also conducted using a finite element method. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by URANS simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.

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