Generic Experimental Investigations of Thermohydraulic Instabilities With Void Fraction Measurement at Natural Circulation Test Facility Geneva

Author(s):  
Tim Cloppenborg ◽  
Christoph Schuster ◽  
Antonio Hurtado

Passive systems like natural circulation (NC) loops can offer reliable and cost efficient alternatives to common active systems for decay heat removal in nuclear power plants. During the transition between stable single and stable two phase flows, instabilities e. g. flashing and geysering may occur in the riser due to low system pressure and saturation temperature conditions. These instabilities may cause severe stress to the system components. This paper presented some results of the study on the decay heat removal system based on natural circulation, performed on the open loop NC test facility GENEVA, built at TU Dresden in 2013. 16 probes were used to determine void fraction along the riser on nine different levels in high time and spatial resolution, and stability maps was created for riser with inner diameters of 20 mm and 38 mm and up to 85 kW evaporator power.

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):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Jörg 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. This can lead to the development of more compact and more efficient components, e.g. heat exchanger and compressors. A specific field of interest is a new 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 station blackout this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. This scenario has already been investigated by means of the thermo-hydraulic code ATHLET, numerically demonstrating the operation of this system for more than 72 h. The practical demonstration is carried out within the Project “sCO2-HeRo”, funded by the European Commission, in which a small scale demonstration unit of the turbo compressor shall be installed at the PWR glass model at GfS, Essen, Germany. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger needs to be as compact and efficient as possible. Therefore, a diffusion welded plate heat exchanger (DWHE) was developed and manufactured at IKE. It has been designed with rectangular mini-channels (0.5–3 mm hydraulic diameter) to ensure high compactness and high heat transfer coefficients. Due to uncertainties the DWHE has to be tested in regard to the actual possible transferrable heat power and to the pressure loss. According to this demand a multipurpose facility has been built at 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 the test section. The test section itself can be exchanged by other ones for various investigations. After the test section, the CO2 pressure is reduced and the liquid is stored in storage tanks, from where it is evaporated and compressed again. 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. The first subject of interest will be the study of the thermal behavior of a DWHE using supercritical CO2 as a working fluid close to its critical point. Experiments concerning pressure loss and heat transfer will be carried out as a start for fundamental investigation of heat transfer in mini-channels. This paper contains a detailed description of the test facility, of the first test section and first results regarding heat transfer power and pressure loss.


Author(s):  
A. K. Nayak ◽  
Mukesh Kumar ◽  
Sumit V. Prasad ◽  
V. Jain ◽  
D. K. Chandraker

Removal of decay heat with nonavailability of active systems is a safety issue especially during station blackout (SBO) in a light water reactor. Passive systems are being incorporated in the new designs of nuclear reactors for this purpose. Some of the advanced reactors such as Indian advanced heavy water reactor (AHWR) have dedicated isolation condensers (ICs) which are submerged in large water pool called gravity driven water pool (GDWP). These ICs remove decay heat from the core by natural circulation cooling and dissipate it to the GDWP by natural convection. There is a concern that cracks may develop in the GDWP if a large seismic event similar to Fukushima type occurs. In that case, the pool water is lost and it can threaten the core coolability because of loss of heat sink. In AHWR, the cracks in the water pool leads to the relocation of the water of the pool to the reactor cavity. Feeders of AHWR are positioned in the reactor cavity. Thus, the water relocated in the cavity, will eventually submerge the feeders and these submerged feeders have the potential to remove the decay heat of the core. However, the feeders are located at a lower elevation as compared to the core, and hence, there is concern on the heat removal capability by the submerged feeders by natural convection. To understand this aspect and to establish the core coolability under the above-mentioned conditions, experiments were performed in a full-scale test facility of AHWR. Experiments showed that the decay heat can be safely removed in natural circulation mode of cooling with heat sink located at lower elevation than the heat source.


2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.


2020 ◽  
Vol 328 ◽  
pp. 01009
Author(s):  
František Világi ◽  
Branislav Knížat ◽  
Róbert Olšiak ◽  
Marek Mlkvik ◽  
Peter Mlynár ◽  
...  

The natural circulation helium loop is an experimental facility designed for the research of the possibility of utilizing natural convection cooling for the case of decay heat removal from a fast nuclear reactor. This concept would bring an improved automated safety system for future nuclear power plants operating a gas-cooled reactor. The article presents a new possibility of direct use of energy conservation laws in a 1D simulation of natural circulation loops. The calculation is performed by a triple iteration process, nested into each other. The results of the calculations showed good agreement with the measurements at steady state. A calculation with the proposed model at unsteady state is not yet possible, especially because of the exclusion of heat accumulation into the material.


2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.


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