Construction of a High Temperature Grade 91 Sodium Component Test Loop

2010 ◽  
Vol 654-656 ◽  
pp. 416-419
Author(s):  
Hyeong Yeon Lee ◽  
Jae Han Lee ◽  
Tae Ho Lee ◽  
Jae Hyuk Eoh Lee ◽  
Tae Joon Kim ◽  
...  

A large scale sodium test facility of ‘CPTL’(Component Performance Test Loop) for simulating thermal hydraulic behavior of the Korean demonstration fast reactor components such as IHX(Intermediate Heat Exchanger), DHX(Decay Heat Removal Heat Exchanger) and sodium pump under development by KAERI is to be constructed. The design temperature of this test loop is 600°C and design pressure is 1MPa. The three heat exchangers are made of Grade 91 steel. Another sodium test facility of the ‘STEF’(Sodium Thermal-Hydraulic Experimental Facility) will be constructed next to the CPTL facility to simulate the passive decay heat removal behavior in the sodium cooled fast reactor. In this paper, the overall facility features of the CPTL and STEF are introduced and preliminary conceptual design of the facilities are carried out.

Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.


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.


2018 ◽  
Vol 68 (1) ◽  
pp. 1-10
Author(s):  
František Dzianik ◽  
Štefan Gužela ◽  
Eva Puškášová

Abstract The paper deals with the process properties in terms of the heat transfer, i.e. the thermal performance of the thermal-process units within a helium loop intended for the testing of the decay heat removal (DHR) from the model of the gas-cooled fast reactor (GFR). The system is characterised by a natural circulation of helium, as a coolant, and assume the steady operating conditions of the circulation. The helium loop consists of four main components: the model of the gas-cooled fast reactor, the model of the heat exchanger for the decay heat removal, hot piping branch and cold piping branch. Using the thermal calculations, the thermal performance of the heat exchanger model and the thermal performance of the gas-cooled fast reactor model are determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.


Author(s):  
V. Vinod ◽  
V. A. Suresh Kumar ◽  
I. B. Noushad ◽  
T. R. Ellappan ◽  
K. K. Rajan ◽  
...  

In pool type Fast Breeder Reactors (FBR) a passive Safety Grade Decay Heat Removal (SGDHR) system removes decay heat produced in the core when normal heat removal path through steam water system is not available. This is essential to maintain the core temperatures within limits. A Decay Heat Exchanger (DHX) picks the heat from the pool and transfers the heat to atmosphere through sodium to Air Heat Exchanger (AHX) situated at high elevation. Due to the temperature differences existent in the system density differences are generated causing a buoyant convective heat transfer. The system is completely passive as primary sodium, secondary sodium and air flows under natural convection. DHX is a sodium to sodium counter flow heat exchanger with primary sodium on shell side and secondary sodium on tube side. AHX is a cross flow heat exchanger with sodium on tube side and air flows in cross flow across the finned tubes. Capacity of a single loop of SGDHR is 8MW. Four such loops are available for the decay heat removal. It has been seen that the decay heat removal to a large extent depends on the AHX performance. AHX tested have shown reduced heat removal capacity much as 30 to 40%, essentially due to the bypassing of the finned tubes by the air. It was felt that a geometrically similar AHX be tested in sodium. Towards this a 2MW Sodium to air heat exchanger (AHX) was tested in the Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Center for Atomic Research (IGCAR), Kalpakkam. The casing arrangement of the AHX was designed to minimise bypassing of air.


Author(s):  
A. K. Nayak ◽  
Mukesh Kumar ◽  
A. K. Vishnoi ◽  
Vikas Jain ◽  
D. K. Chandraker

Decay heat removal for prolonged period of station blackout (SBO) is a safety concern of the nuclear reactors. Aftermath of Fukushima, safety evaluation (performance under severe conditions: stress test) of the reactors was carried out worldwide. It includes establishment of grace period of the reactors. Similar exercises for advanced heavy water reactor (AHWR) were also performed and the design of AHWR was established for its robustness against such events. Decay heat removal during extended SBO is such a condition to be qualified. In this regard, experiments in the integral test loop (ITL), a full scale test facility of AHWR, were conducted for continuous 7 days of extended SBO. Experiment was started with 6.8 MPa as the initial reactor pressure and decay heat removal was demonstrated for 7 days of SBO by passive means. It is observed that the pressure falls down to 1 MPa in 3 h. The design of AHWR was evaluated from safety critical aspects during such an event experimentally. During this event, the clad surface temperature was found to be well within safe limits of operations. As a result of this experiment, it can be concluded that the design of AHWR is capable to remove decay heat for 7 days of SBO with sufficient safety margins.


Author(s):  
Kwi Lim Lee ◽  
Kwi Seok Ha ◽  
Hae Yong Jeong ◽  
Won Pyo Chang

Korea Atomic Energy Research Institute (KAERI) has been developing a conceptual design of the demonstration fast reactor (DFR), which is the pool type sodium cooled fast reactor with the thermal power of 1548.2 MW and the core loaded with metal fuel. The DFR is composed of a Primary Heat Transport System (PHTS), an Intermediate Heat Transport System (IHTS), a Steam Generating System (SGs) and a decay heat removal system (DHRS). The DHRS is composed of 2 units of Passive Decay-heat Removal Circuits (PDRC) and 2 units of Active Decay-heat Removal Circuits (ADRC). The PDRC consists of two independent loops with sodium-to-sodium Decay Heat eXchanger (DHX) and natural-draft sodium-to-Air Heat eXchanger (AHX). The ADRC consists of two independent loops with sodium-to-sodium DHX and Forced-Draft sodium-to-air Heat eXchanger (FDHX) located in the upper region of the reactor building. The PDRC is very different from that of KALIMER-600 on the points of the submerged location and the heat transfer mechanism. For the identification of safety characteristics, 5 DBE’s (Design Bases Events) are analyzed using the MARS-LMR code. The representative DBE’s are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Reactor Vessel Leak and Pipe Break. As a result, it is identified that the DFR were appropriately performed as designed and the temperatures of the fuel and the structure were evaluated to satisfy the criteria.


Author(s):  
Nina Yue ◽  
Rong Cai ◽  
Yun Wang ◽  
Suizheng Qiu ◽  
Dalin Zhang

A sodium-cooled fast reactor is a significant candidate for future power reactor systems. Decay heat removal is an essential function of reactor safety systems The decay heat removal system should have the capacity to remove the decay heat with natural circulation in any accident. There are three types of decay heat removal systems, namely direct reactor auxiliary cooling system, primary reactor auxiliary cooling system, and intermediate reactor auxiliary cooling system. The one dimensional systems analysis code THACS was applied to conduct transient analyses of a sodium-cooled fast reactor, and the capabilities of three types of decay heat removal systems against a station blackout accident were compared. The results indicate that these three types of decay heat removal systems can remove the residual heat effectively. For large-scale sodium-cooled fast reactor, the capabilities of primary reactor auxiliary cooling system and intermediate reactor auxiliary cooling system were better, because the cold sodium from the penetrating heat exchanger in these two auxiliary cooling systems could directly flow into the core assemblies.


Author(s):  
Xiaokun Wang ◽  
Donghui Zhang

CEFR (China Experimental Fast Reactor) is a typical pool type SFR (Sodium cooled Fast Reactor) which relies on independent positive DHRS (Decay Heat Removal System). This paper describes a new 1-D transient thermal hydraulic code named LR which is based on DHRS of CEFR. LR code can be used for coupling calculations of heat exchange and flow in the primary side of DHX (decay heat exchanger), middle loop of DHRS and air side of AHX (air heat exchanger). The calculations of these three loops are all based on 1-D model for the consideration of the structure of this system, the local heat transfer coefficients and the drag coefficients. Meanwhile, boundary conditions include sodium pool temperature field around the DHX, air temperature field around the AHX and air stack, as well as the condition of the air inlet on AHX. And the initial condition is the steady state which is calculated from the boundary conditions at the initial time. We have simulated the transient process of DHRS in CEFR with this code, the result of which is closed to that of Russian CBTO and fits the conclusions of theoretical analysis.


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.


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