Review and Status of the Gen-IV CANDU-SCWR Passive Moderator Core Cooling System

2008 ◽  
Author(s):  
H. F. Khartabil
Keyword(s):  
Heat Sink ◽  
Cooling System ◽  
Natural Circulation ◽  
Power Level ◽  
Heat Removal ◽  
Circulation System ◽  
The Past ◽  
Test Loop ◽  
Flow Oscillations ◽  
Core Cooling

Enhanced safety is an important priority in the development of Generation IV reactors, which can be accomplished through the use of improved passive heat removal systems. In CANDU® reactors, the separation between the low-pressure moderator and high-pressure coolant provides a unique passive heat sink for decay heat removal during accident scenarios. Methods for enhancing this passive heat sink for the GenIV CANDU-SCWR (supercritical water cooled reactor) have been under investigation for the past several years to support a “no core melt” reactor design concept (1, 2). Initially, to test feasibility, tests and analysis at AECL studied a full-height passive cooling loop and showed that a flashing-driven natural circulation system was possible in principle. However, flow oscillations were observed at low powers and could not be readily explained through analysis. While these oscillations were not considered to be detrimental to the heat removal capability, additional separate-effects experiments were conducted and causal mechanisms proposed for the oscillations. In addition, these separate effects tests suggested that oscillations could be avoided at any power level by suitable design. A new test loop with a more representative geometry was recently constructed and commissioned. Preliminary commissioning tests confirmed conclusions from the separate effects tests. In this paper, the new tests are compared to the past tests to explain the improved and more stable loop operation. This comparison suggests that a complete system coupled to an ultimate heat sink has the potential to improve loop operation even more by eliminating or significantly reducing flow oscillations at low powers. Plans for validating this conclusion will be provided.

The Plant Feature and Performance of DMS (Double MS: Modular Simplified and Medium Small Reactor)

2008 ◽  
Author(s):  
Yukiko Kawabata ◽  
Masayoshi Matsuura ◽  
Shizuka Hirako ◽  
Takashi Hoshi
Keyword(s):  
Economic Efficiency ◽  
Cooling System ◽  
Natural Circulation ◽  
Circulation System ◽  
Surface Separation ◽  
Power Company ◽  
Loss Of Coolant Accident ◽  
Core Cover ◽  
And Performance ◽  
Core Cooling

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.

Preliminary Analysis of Unprotected Loss of Heat Sink in Small Long Life Gas Cooled Fast Reactor

2015 ◽  
Vol 751 ◽  
pp. 268-272
Author(s):  
Su'ud Zaki ◽  
Nuri Trianti ◽  
Rosidah M. Indah
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Channel Model ◽  
Cooling System ◽  
Hot Spot ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reducing Power ◽  
Secondary System ◽  
Heat Removal System

The failure of the secondary side in Gas Cooled Fast Reactor system, which may contain co-generation system, will cause loss of heat sink (LOHS) accident. In this study accident analysis of unprotected loss of heat sink due to the failure of the secondary cooling system has been investigated. The thermal hydraulic model include transient hot spot channel model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is also plays important role to limit the level of consequence during the accident. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate the failure of the secondary system by reducing power through reactivity feedback and remove the rest of heat through natural circulations based decay heat removal (RVACS system).

Study of the Behaviors of an Open-Loop Heat-Removal System

2018 ◽  
Author(s):  
Xianmao Wang ◽  
Yonggang Shen ◽  
Jiang Yang ◽  
Yong Ouyang ◽  
Min Rui ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Calculation Model ◽  
Open Loop ◽  
Cooling Systems ◽  
Core Cooling ◽  
Passive Cooling Systems ◽  
Heat Removal System ◽  
Removal System

In the third generation of nuclear reactors, passive systems have been widely used such as passive core cooling system and passive containment cooling system, which usually relay on natural circulation induced by buoyancy force to remove heat. Most of these passive cooling systems are closed-loop natural circulations. In recent years, some open-loop heat-removal systems have also been put forward. Open-loop heat-removal systems have its own advantages such as its simplification and low costs. However, the thermal-hydraulic behaviors of open-loop heat-removal systems are still not totally clear and need further study. In this study, a simplified open-loop passive containment cooling system is studied. A calculation model is built based on RELAP SCDAPSIM code. The thermal-hydraulic behaviors of the system are studied. By changing some key parameters of the system, the influences of these parameters on the system are evaluated.

Experimental investigation on instability characteristics of loss of heat sink accident in a natural circulation system

2021 ◽  
Vol 155 ◽  
pp. 108143
Author(s):  
Yuqi Lin ◽  
Puzhen Gao ◽  
Xianbing Chen ◽  
Solomon Bello ◽  
Chunping Tian ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Heat Sink ◽  
Natural Circulation ◽  
Circulation System

Description of Typical Flow Instabilities in the Natural Circulation System

2002 ◽  
Author(s):  
Shengyao Jiang ◽  
Xingtuan Yang ◽  
Youjie Zhang
Keyword(s):  
Natural Circulation ◽  
Great Influence ◽  
Operating Conditions ◽  
Boiling Water ◽  
Flow Instabilities ◽  
Circulation System ◽  
Boiling Water Reactor ◽  
Primary Loop ◽  
Water Reactor ◽  
Test Loop

The experiments were performed on the test loop HRTL-5, which simulates geometry and system design of the 5-MW Nuclear Heating Reactor developed by the Institute of Nuclear Energy Technology, Tsinghua University. Because of the difference of the geometry design and operating conditions between the heating reactor and the boiling water reactor, the flow behavior presents great differences too, some of which haven’t been deeply studied so far. Results show that in heating reactor, sub-cooled boiling, condensation and flashing play an important role on the flow instabilities of the natural circulation system. Correspondingly, geysering instability, flashing instability, and flow excursion are the very typical instabilities occurring in the primary loop of HRTL-5, which are different from those in boiling water reactor conditions. The compressibility of the steam space on the top of the primary loop has also great influence on the instability of the natural circulation system.

Performance of Sodium Natural Circulation Loop for Passvie Decay Heat Removal

2008 ◽  
Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  
Keyword(s):  
Fast Reactor ◽  
System Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Circulation Loop ◽  
Ambient Atmosphere ◽  
Steady State Condition ◽  
Decay Heat ◽  
Natural Circulation Loop ◽  
Test Loop

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.

Numerical Analysis of Natural Circulation Phenomena of Supercritical Fluids

2013 ◽  
Author(s):  
Jeffrey Samuel ◽  
Glenn Harvel ◽  
Igor Pioro
Keyword(s):  
Scaling Laws ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Natural Circulation ◽  
Operating Conditions ◽  
Passive Safety System ◽  
Test Loop ◽  
R 134A ◽  
Carbon Dioxide Co2 ◽  
Core Cooling

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.

Effects of loss of heat sink transient on flow characteristics in a closed natural circulation system

2021 ◽  
Vol 381 ◽  
pp. 111331
Author(s):  
Solomon Bello ◽  
Puzhen Gao ◽  
Samuel Abiodun Olatubosun ◽  
Yuqi Lin
Keyword(s):  
Heat Sink ◽  
Natural Circulation ◽  
Flow Characteristics ◽  
Circulation System

scholarly journals Experimental evaluation of natural convection in the IPR-R1 Triga research reactor at 264 kW and 105 kW

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Amir Zacarias Mesquita
Keyword(s):  
Natural Convection ◽  
Technology Development ◽  
Cooling System ◽  
Natural Circulation ◽  
Fuel Temperature ◽  
Operational Parameters ◽  
Reactor Operating ◽  
Core Cooling ◽  
Reactor Pool ◽  
Forced Cooling

In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 Triga Reactor of the Nuclear Technology Development Center (CDTN), experimental measurements were done with the reactor operating at power levels of 265 kW and 105 kW, with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity, and influence of cooling system. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 online performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database. This paper summarizes the behavior of some operational parameters, and in particular, the evolution of the temperature in the fuel element centerline positioned in the core hottest location. The natural circulation test was performed to confirm the cooling capability of the natural convection in the IPR-R1 reactor. It was confirmed that the IPR-R1 has capability of long-term core cooling by natural circulation operating at 265 kW. The measured maximum fuel temperature of about 300 oC was lower than the operating limit of 550 oC. It has been proven that without cooling in the primary the gamma dose rate above reactor pool at high power levels was rather high.

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