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

2021 ◽  
Author(s):  
Masayoshi Mori ◽  
Kosuke Ono ◽  
Tetsuya Takada ◽  
Yasunori Yamamoto
Keyword(s):  
Experimental Study ◽  
Cooling System ◽  
Natural Circulation ◽  
Circulation Flow ◽  
Core Cooling

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

2020 ◽  
Author(s):  
Yasunori Yamamoto ◽  
Masayoshi Mori ◽  
Kosuke Ono ◽  
Tetsuya Takada
Keyword(s):  
Steady State ◽  
Cooling System ◽  
Natural Circulation ◽  
Operating Conditions ◽  
Quasi Steady State ◽  
Circulation Flow ◽  
Term Operation ◽  
Accident Conditions ◽  
Core Cooling

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.

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.

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.

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.

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.

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