Assessment of the MARS code for design of water pool-type reactor cavity cooling system in HTGR

2011 ◽  
Vol 241 (5) ◽  
pp. 1715-1725 ◽  
Author(s):  
Yun-Je Cho ◽  
Goon-Cherl Park
Keyword(s):  
Cooling System ◽  
Water Pool ◽  
Type Reactor ◽  
Cavity Cooling ◽  
Pool Type

Experiments on a Water Pool-Type Reactor Cavity Cooling System in a High-Temperature Gas-Cooled Reactor

2007 ◽  
Vol 159 (1) ◽  
pp. 39-58 ◽  
Author(s):  
Hyoung Kyu Cho ◽  
Yun Je Cho ◽  
Moon Oh Kim ◽  
Goon Cherl Park
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Water Pool ◽  
Type Reactor ◽  
Cavity Cooling ◽  
Pool Type ◽  
High Temperature Gas

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.

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

Two-phase flow measurements and observations in a cooling panel of the reactor cavity cooling system

2021 ◽  
Vol 131 ◽  
pp. 103578
Author(s):  
R. Vaghetto ◽  
S. Yang ◽  
D. Hodge ◽  
Y. Hassan
Keyword(s):  
Two Phase Flow ◽  
Cooling System ◽  
Phase Flow ◽  
Two Phase ◽  
Flow Measurements ◽  
Cavity Cooling

scholarly journals Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

2014 ◽  
Vol 2014 ◽  
pp. 1-14 ◽  
Author(s):  
Hyun-Sik Park ◽  
Byung-Yeon Min ◽  
Youn-Gyu Jung ◽  
Yong-Cheol Shin ◽  
Yung-Joo Ko ◽  
...  
Keyword(s):  
System Analysis ◽  
Cooling System ◽  
Heat Removal ◽  
Test Facility ◽  
Integral Type ◽  
Type Reactor ◽  
Integral Effect ◽  
Post Test ◽  
Design Characteristics ◽  
And Performance

To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribute to the SMART design licensing by providing proper test data for validating the system analysis codes. A typical scenario of SBLOCA was analyzed using the MARS-KS code to assess the thermal-hydraulic similarity between the SMART design and the VISTA-ITL facility, and a posttest simulation on a SBLOCA test for the shutdown cooling system line break has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code.

Scaling Analysis of Reactor Cavity Cooling System in HTR

2019 ◽  
Vol 206 (4) ◽  
pp. 527-543
Author(s):  
Thiago D. Roberto ◽  
Celso M. F. Lapa ◽  
Antonio C. M. Alvim
Keyword(s):  
Cooling System ◽  
Scaling Analysis ◽  
Cavity Cooling
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