Experimental Study of the Effect of Hydrogen Inflow on Passive Core Cooling System With Natural Circulation Flow

Author(s):  
Yasunori Yamamoto ◽  
Masayoshi Mori ◽  
Kosuke Ono ◽  
Tetsuya Takada

Abstract Isolation Condenser (IC) is one of the passive core cooling systems with natural circulation flow, which is effective for safety measures against station black out. Once core uncover occurs, hydrogen generated in the core affects operating condition of ICs. To use ICs as an important safety measure not only for transient conditions but also for accident conditions, robustness of ICs against hydrogen inflow must be understood well. In this study, experiments with high pressure steam were conducted using experimental setup simulating IC, where helium was injected to simulate hydrogen effects. When the pressure in an accumulator increased high enough, natural circulation flow generated in the experimental loop. After the long-term operation, the pressure and the natural circulation flow rate achieved nearly constant. The pressure at quasi-steady state increased with increasing the helium injection amount. The pressure difference in a section including outlet side of a vertical pipe was slightly increased when helium was injected which may have indicated that the helium accumulated in the section and caused increment of the pressure loss. The startup pressure of the IC simulator also increased when helium was injected, where the driving force by the water head difference also decreased. Though long-term operations were performed after helium injection, the effect of injected helium on operating conditions of the IC remained for quasi-steady state conditions.

2016 ◽  
Vol 2016 ◽  
pp. 1-12 ◽  
Author(s):  
Pengcheng Zhao ◽  
Kangli Shi ◽  
Shuzhou Li ◽  
Jingchao Feng ◽  
Hongli Chen

Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.


2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.


2011 ◽  
Vol 63 (9) ◽  
pp. 1906-1912 ◽  
Author(s):  
Simos Malamis ◽  
Andreas Andreadakis ◽  
Daniel Mamais ◽  
Constantinos Noutsopoulos

The aim of this work was to evaluate the long-term performance of a Membrane Bioreactor (MBR) that operated continuously for 2.5 years and to assess membrane fouling and biomass activity under various operating conditions. Furthermore, a method for the characterisation of influent wastewater was developed based on its separation into various fractions. The MBR system operated at the solids retention times (SRT) of 10, 15, 20 and 33 days. The increase of SRT resulted in a decrease of the fouling rate associated with the reduction of extracellular polymeric substances. Moreover, the SRT increase resulted in a significant reduction of the Oxygen Uptake Rate (OUR) due to the lower availability of substrate and in a notable decrease of the maximum OUR since high SRT allowed the development of slower growing microorganisms. Biomass consisted of small flocs due to extensive deflocculation caused by intense aeration. Finally, the method developed for wastewater characterisation is straightforward and less time consuming than the usual method that is employed.


2020 ◽  
Vol 14 (1) ◽  
pp. 55-60
Author(s):  
O. N. Didmanidze ◽  
R. T. Khakimov ◽  
E. P. Parlyuk ◽  
N. A. Bol’shakov

Global car manufacturers wish to increase the number of manufactured products, reduce their cost and labor input. The choice of research areas, design and technological developments in radiator construction is an extremely important and urgent task, due to the mass production of radiators for tractors and automobiles on the one hand, and the favorable development prospects of these interrelated industries, on the other. (Research purpose) To substantiate theoretically and experimentally the use of a combined cooling system containing both aluminum and polymeric water radiators and similarly liquid-oil heat exchangers based on the four principles listed above on automobiles and tractors. (Materials and methods) The authors performed bench tests using a special wind tunnel to study the thermal and aerodynamic characteristics of a prototype tractor radiator with a polyurethane core. After reaching the steady-state operating mode of the installation, the experimental values were determined for the control and measuring instruments. (Results and discussion) The authors carried out measurements of all parameters of both coolants in series at each steady-state operating mode of the bench. They obtained the main indicators dependences (reduced heat transfer, aerodynamic and hydraulic drag) of the heat exchanger, close to the operating conditions of the vehicles. (Conclusions) A prototype MTZ-80 radiator with a polyurethane core has great prospects as a future alternative radiator. An increase by 10-15 percent in the radiator heat transfer is possible by using aluminum fi ns on the surface of the polyurethane plate. A 15-20 percent reduction in hydrodynamic resistance is achieved by increasing the diameter of the capillary throughput in a polyurethane plate and the number of plates themselves in the radiator cell.


2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5


Author(s):  
Yukiko Kawabata ◽  
Masayoshi Matsuura ◽  
Shizuka Hirako ◽  
Takashi Hoshi

The Japan Atomic Power Company has initiative in developing the DMS concept as a 400MWe-class light water reactor. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts and effective use of proven technology. The decrease in primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about two meters compared with 3.7 meters in the conventional BWR. The short active fuel length reduces the drop in core pressure, and overcomes the natural circulation system. And by using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to improve safety efficiency, developing an Emergency Core Cooling System (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following. 1: Eliminating high-pressure injection systems. 2: Adopting passive safety-related systems. 3: Optimizing distribution for the systems and power source for the ECCS. In this way the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of Loss of Coolant Accident (LOCA) analysis, core cover can be achieved by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.


Author(s):  
Jeffrey Samuel ◽  
Glenn Harvel ◽  
Igor Pioro

The feasibility of operating with natural circulation as the normal mode of core cooling has been successfully demonstrated for a few small sized nuclear reactors. Natural circulation is being considered for cooling the core of a nuclear reactor under normal operating conditions in several advanced reactor concepts being developed today. Although studies have been conducted in natural circulation for many decades, using natural circulation as the primary cooling mechanism for nuclear reactors or as a passive safety system requires a comprehensive understanding of local and integral system phenomena, validated benchmark data, accurate predictive tools, and reliability analysis methods. As full-scale experiments of supercritical water are expensive, scaling laws can be applied to develop test matrices using modelling fluids to reproduce similar conditions in a scaled-down experimental thermalhydraulic loop. The main aim of this work is to understand the natural circulation phenomena by analyzing water and modelling fluids such as Carbon dioxide (CO2) and Freon 134a (R-134a). The use of the modelling fluids at subcritical, pseudocritical and supercritical pressures is discussed along with fluid-to-fluid scaling techniques. The results from a one-dimensional numerical model developed using MATLAB to calculate the steady-state mass flow rate and heat transport characteristics of an experimental natural circulation test loop are presented and analyzed.


2010 ◽  
Author(s):  
H. S. Yang ◽  
D. L. Kim ◽  
S. H. Sohn ◽  
J. H. Lim ◽  
Y. S. Choi ◽  
...  

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