Numerical Analysis of Single Phase Thermal Stratification in both Cold Legs and Downcomer by Emergency Core Cooling System Injection : A Study on the Necessity to Consider Buoyancy Force Term

2017 ◽  
Vol 29 (12) ◽  
pp. 654-662
Author(s):  
Gong Hee Lee ◽  
Ae Ju Cheong
Keyword(s):  
Numerical Analysis ◽  
Thermal Stratification ◽  
Buoyancy Force ◽  
Single Phase ◽  
Cooling System ◽  
Force Term ◽  
Core Cooling

scholarly journals Suggestion for the Proper Treatment of Buoyancy Force Effect on Single Phase Thermal Stratification Phenomenon

2019 ◽  
Vol 128 ◽  
pp. 10009
Author(s):  
Gong-Hee Lee
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Single Phase ◽  
Cooling System ◽  
Turbulent Transport ◽  
Governing Equations ◽  
Service Testing ◽  
Ansys Cfx ◽  
Core Cooling ◽  
Coolant System

In the operating nuclear power plant, piping and components installed in the in-service testing (IST) systems connected to the reactor coolant system are potentially exposed to fatigue caused by thermal stratification flow, possibly threatening its integrity. The emergency core cooling system (ECCS) is one of candidates for which the thermal stratification phenomenon can occur. During the ECCS operating period, buoyancy due to density difference may have significant influence on thermal-hydraulic characteristics of the mixing flow. Therefore, in this study, the need to consider proper buoyancy models in governing equations, especially turbulent transport equations, for accurate prediction of single phase thermal stratification by ECCS injection was numerically studied using ANSYS CFX R17.2.

Numerical Investigation of Thermal Stratification in the Primary Circuit of VVER-440 Type Reactors

2006 ◽  
Author(s):  
Ildiko´ Boros ◽  
Attila Aszo´di ◽  
Ga´bor Le´gra´di
Keyword(s):  
Thermal Stratification ◽  
Cooling System ◽  
Maximum Temperature ◽  
Primary Circuit ◽  
Operational Experience ◽  
Extensive Evaluation ◽  
Transient Simulations ◽  
Surge Line ◽  
Core Cooling

Thermal stratification in the primary loops and in the connected pipes can limit the lifetime of the piping, or lead to penetrating cracks due to the stresses caused by the temperature differences and the cyclic temperature changes. Therefore it is essential to determine the thermal hydraulic parameters of the stratified flow. The determination of the affected pipes can be based on the international operational experience and on engineering consideration. The most affected pipes in PWRs are the pressurizer surge line, the injection pipe of the emergency core cooling systems and the feedwater injection pipe of the steam generators. CFD codes can provide an appropriate tool for the examination of the development and the breaking up of the stratification and the determination of the temperature distribution. However, the challenge of the uncertainty of the boundary conditions has to be faced because of the unknown flow circumstances. According to an extensive evaluation, performed in 1998 by the VEIKI, in the VVER-440/213 units of Paks NPP the most affected pipe is the pressurizer surge line [1]. To find out the possible thermal stratification in the surge line, a temperature monitoring system was installed on the YP20 leg of the surge line of the Unit 1 of the Paks NPP in 2000. The measurements showed that during the heat-up period there is a thermal stratification almost all time in the surge line [2]. The maximum temperature differences reach 140 K (140 °C). The surge line has been modeled with the CFD code CFX-5.7. The performed transient simulations confirmed the existence of a thermal stratification in the surge line, but showed permanent recirculation of colder coolant in the lower layer, caused by the asymmetric arrangement of the surge line legs and the asymmetric connection of the two legs to the main loop. In this paper, the surge line model and the results of the transient simulations are presented. The CFD model of the injection pipe of the high pressure Emergency Core Cooling System and the performed simulations for the analysis of occurrence of thermal stratification are presented as well.

Development of Passive In-Core Cooling System for Nuclear Safety Using Hybrid Heat Pipe

2016 ◽  
Vol 196 (3) ◽  
pp. 598-613 ◽  
Author(s):  
Kyung Mo Kim ◽  
Yeong Shin Jeong ◽  
In Guk Kim ◽  
In Cheol Bang
Keyword(s):  
Heat Pipe ◽  
Cooling System ◽  
Nuclear Safety ◽  
Core Cooling

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

Exploring the Use of Alumina Nanofluid as Emergency Coolant for Nuclear Fuel Bundle

2018 ◽  
Vol 11 (2) ◽  
Author(s):  
A. S. Chinchole ◽  
Arnab Dasgupta ◽  
P. P. Kulkarni ◽  
D. K. Chandraker ◽  
A. K. Nayak
Keyword(s):  
Heat Transfer ◽  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Cooling System ◽  
Nuclear Reactors ◽  
Electrical Heating ◽  
High Dose ◽  
Potential Candidate ◽  
Alumina Nanofluid ◽  
Core Cooling

Abstract Nanofluids are suspensions of nanosized particles in any base fluid that show significant enhancement of their heat transfer properties at modest nanoparticle concentrations. Due to enhanced thermal properties at low nanoparticle concentration, it is a potential candidate for utilization in nuclear heat transfer applications. In the last decade, there have been few studies which indicate possible advantages of using nanofluids as a coolant in nuclear reactors during normal as well as accidental conditions. In continuation with these studies, the utilization of nanofluids as a viable candidate for emergency core cooling in nuclear reactors is explored in this paper by carrying out experiments in a scaled facility. The experiments carried out mainly focus on quenching behavior of a simulated nuclear fuel rod bundle by using 1% Alumina nanofluid as a coolant in emergency core cooling system (ECCS). In addition, its performance is compared with water. In the experiments, nuclear decay heat (from 1.5% to 2.6% reactor full power) is simulated through electrical heating. The present experiments show that, from heat transfer point of view, alumina nanofluids have a definite advantage over water as coolant for ECCS. Additionally, to assess the suitability of using nanofluids in reactors, their stability was investigated in radiation field. Our tests showed good stability even after very high dose of radiation, indicating the feasibility of their possible use in nuclear reactor heat transfer systems.

A Numerical Analysis on the System Impedance in a Fan Cooling System

2003 ◽  
Author(s):  
Dong-Il Kim ◽  
Ki-So Bok ◽  
Han-Bae Lee
Keyword(s):  
Numerical Analysis ◽  
Pressure Drop ◽  
Flow Rate ◽  
Pressure Difference ◽  
Cooling System ◽  
Computational Time ◽  
Computational Domain ◽  
Impedance Curve ◽  
Fan Cooling ◽  
Large Investment

To seek the fan operating point on a cooling system with fans, it is very important to determine the system impedance curve and it has been usually examined with the fan tester based on ASHRAE standard and AMCA standard. This leads to a large investment in time and cost, because it could not be executed until the system is made actually. Therefore it is necessary to predict the system impedance curve through numerical analysis so that we could reduce the measurement time and effort. This paper presents how the system impedance curve (pressure drop curve) is computed by CFD in substitute for experiment. In reverse order to the experimental principle of the fan tester, pressure difference was adopted first as inlet and outlet boundary conditions of the system and then flow rate was calculated. After determining the system impedance curve, it was compared with experimental results. Also the computational domain of the system was investigated to minimize computational time.

The AP1000 LBLOCA Analysis Using SCDAP/RELAP5

2014 ◽  
Author(s):  
Heng Xie
Keyword(s):  
Heat Conduction ◽  
Simulation Model ◽  
Cooling System ◽  
Branch Pipe ◽  
Fluid System ◽  
Performance Simulation ◽  
Fuel Rod ◽  
Axial Heat ◽  
Core Cooling ◽  
Coolant System

In this study, a transient performance simulation model of AP1000 using SCDAP/RELAP5 4.0 is developed. The reactor coolant system (RCS) and passive core cooling system (PXS) are modeled respectively. Various kinds of hydrodynamic component including Volume, Junction, Separator, Accumulator, Branch, Pipe, Valve and Pump are adopted to simulate the fluid system of AP1000. The DECLG (double-end rupture of cold leg) accident is simulated and analyzed. To study the effect of axial heat conduction, two kinds of heat structure with and without reflooding model are employed to simulate the fuel rod respectively. The comparison shows that the 2D heat conduction play important role in the reflooding process.

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