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).

The Choose of Decay Heat Removal Systems of SFR

2018 ◽  
Author(s):  
Nina Yue ◽  
Rong Cai ◽  
Yun Wang ◽  
Suizheng Qiu ◽  
Dalin Zhang
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Decay Heat ◽  
Station Blackout ◽  
The One ◽  
Safety Systems ◽  
Heat Removal System

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.

Unprotected Loss of Flow Accident in Small Long Life Gas Cooled Fast Reactor

2015 ◽  
Vol 751 ◽  
pp. 263-267
Author(s):  
Su'ud Zaki
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reducing Power ◽  
Inherent Safety ◽  
Fast Reactors ◽  
Transient Model ◽  
Long Life ◽  
Heat Removal System

In post Fukushima nuclear accidents inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Gas cooled fast reactors is one of the important candidate of 4th generation nuclear power plant and in this paper the safety analysis related to unprotected loss of flow in small long life gas cooled fast reactors has been performed. Accident analysis of unprotected loss of flow include coupled neutronic and thermal hydraulic analysis which include adiabatic model in nodal approach of time dependent multigroup diffusion equations. The thermal hydraulic model include transient 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 investigated. 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 complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations.

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

2015 ◽  
Vol 52 (9) ◽  
pp. 1102-1121 ◽  
Author(s):  
Osamu Watanabe ◽  
Kazuhiro Oyama ◽  
Junji Endo ◽  
Norihiro Doda ◽  
Ayako Ono ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Evaluation Methodology ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

scholarly journals CFD Analysis of the Passive Decay Heat Removal System of an LBE-Cooled Fast Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Jiarun Mao ◽  
Lei Song ◽  
Yuhao Liu ◽  
Jiming Lin ◽  
Shanfang Huang ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Engineering Project ◽  
Decay Heat ◽  
The Core ◽  
Main Vessel ◽  
Heat Removal System ◽  
Removal System

This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

scholarly journals An experimental study on natural circulation decay heat removal system for a loop type fast reactor

2016 ◽  
Vol 53 (9) ◽  
pp. 1385-1396 ◽  
Author(s):  
Ayako Ono ◽  
Hideki Kamide ◽  
Jun Kobayashi ◽  
Norihiro Doda ◽  
Osamu Watanabe
Keyword(s):  
Experimental Study ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

scholarly journals Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

2012 ◽  
Vol 2012 ◽  
pp. 1-14 ◽  
Author(s):  
H. Yamano ◽  
S. Kubo ◽  
Y. Shimakawa ◽  
K. Fujita ◽  
T. Suzuki ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Natural Circulation ◽  
Heat Removal ◽  
Severe Accident ◽  
Safety Requirements ◽  
Generation Plant ◽  
Safety Design ◽  
Phenomenological Consideration ◽  
Heat Removal System

As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR) adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS) in design extension conditions (DECs). A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs) and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

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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