Efficacy of air heat exchanger based decay heat removal system through passive mode: A numerical study

2020 ◽  
Vol 28 ◽  
pp. 2286-2294
Author(s):  
D.N. Elton ◽  
U.C. Arunachala ◽  
V.F. Dolfred
Author(s):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Joerg Starflinger

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.


Author(s):  
Dehee Kim ◽  
Jaehyuk Eoh ◽  
Tae-Ho Lee

Sodium-cooled Fast Reactor (SFR) is one of the generation IV (Gen-IV) nuclear reactors. Prototype Gen-IV SFR (PGSFR) is a SFR being developed in Korea Atomic Energy Research Institute (KAERI). Decay Heat Removal System (DHRS) in the PGSFR has a safety function to make shutdown the reactor under abnormal plant conditions. Single DHRS loop consists of sodium-to-sodium decay heat exchanger (DHX), helical-tube sodium-to-air heat exchanger (AHX) or finned-tube sodium-to-air heat exchanger (FHX), loop piping, and expansion vessel. The DHXs are located in the cold pool and the AHXs and FHXs are installed in the upper region of the reactor building. The DHRS loop is a closed loop and liquid sodium coolant circulates inside the loop by natural circulation head for passive system and by forced circulation head for active system. There are three independent heat transport paths in the DHRS, i.e., the DHX shell-side sodium flow path, the DHRS sodium loop path through the piping, the AHX shell-side air flow path. To design the components of the DHRS and to determine its configuration, key design parameters such as mass flow rates in each path, inlet/outlet temperatures of primary and secondary flow sides of each heat exchanger should be determined reflecting on the coupled heat transfer mechanism over the heat transfer paths. The number of design parameters is larger than that of the governing equations and optimization approach is required for compact design of the DHRS. Therefore, a genetic algorithm has been implemented to decide the optimal design point. The one-dimensional system design code which can predict heat transfer rates and pressure losses through the heat exchangers and piping calculates the objective function and the genetic algorithm code searches a global optimal point. In this paper, we present a design methodology of the DHRS, for which we have developed a system code coupling a one-dimensional system code with a genetic algorithm code. As a design result, the DHRS layouts and the sizing of the heat exchangers have been shown.


2021 ◽  
Vol 378 ◽  
pp. 111259
Author(s):  
A. Pantano ◽  
P. Gauthe ◽  
M. Errigo ◽  
P. Sciora

1978 ◽  
Author(s):  
J. E. Kelly ◽  
R. C. Erdmann

Author(s):  
Seong Kuk Cho ◽  
Jekyoung Lee ◽  
Jeong Ik Lee ◽  
Jae Eun Cha

A Sodium-cooled Fast Reactor (SFR) has receiving attention as one of the promising next generation nuclear reactors because it can recycle the spent nuclear fuel produced from the current commercial nuclear reactors and accomplish higher thermal efficiency than the current commercial nuclear reactors. However, after shutdown of the nuclear reactor core, the accumulated fission products of the SFR also decay and release heat via radiation within the reactor. To remove this residual heat, a decay heat removal system (DHRS) with supercritical CO2 (S-CO2) as the working fluid is suggested with a turbocharger system which achieves passive operational capability. However, for designing this system an improved S-CO2 turbine design methodology should be suggested because the existing methodology for designing the S-CO2 Brayton cycle has focused only on the compressor design near the critical point. To develop a S-CO2 turbine design methodology, the non-dimensional number based design and the 1D mean line design method were modified and suggested. The design methodology was implemented into the developed code and the code results were compared with existing turbine experimental data. The data were collected under air and S-CO2 environment. The developed code in this research showed a reasonable agreement with the experimental data. Finally using the design code, the turbocharger design for the suggested DHRS and prediction of the off design performance were carried out. As further works, more effort will be put it to expand the S-CO2 turbine test data for validating the design code and methodology.


2019 ◽  
Author(s):  
Susyadi ◽  
Andi S. Ekariansyah ◽  
Hendro Tjahjono ◽  
D. T. Sony Tjahyani

2016 ◽  
Vol 305 ◽  
pp. 168-178 ◽  
Author(s):  
Fabio Giannetti ◽  
Damiano Vitale Di Maio ◽  
Antonio Naviglio ◽  
Gianfranco Caruso

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