Experiments and Analytical Simulation on Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant

2008 ◽  
Author(s):  
Shuichi Ohmori ◽  
Tadashi Narabayashi ◽  
Michitsugu Mori ◽  
Fumitoshi Watanabe
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
High Pressures ◽  
High Efficiency ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Core Cooling

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. We are developing an innovative idea by applying SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains discharge pressure higher than the supply steam pressure to inject water into the reactor by operating the SI, by supplying water from a pool in a containment vessel and the steam from a reactor pressure vessel (RPV). The SI, passive equipment, is used to replace large rotating machines such as pumps and motors, eliminating the failure probabilities of such active equipment. When the water and steam supply valves open, the SI-driven PCIS (SI-PCIS) will automatically start to inject water into the core to keep the core covered with water. The SI-PCIS works for the range of steam pressure conditions from atmosphere pressure through high pressures, in which the analytical simulations of SI were carried out based on the plenty amount of experimental data using reduced scale SI. We further simulated and evaluated the core cooling and water injection performance of SI-PCIS in BWR using RETRAN-3D code for the case of small LOCA. A reactor, such as ESBWR, equipped with the passive safety system by gravity-driven cooling system (GDCS) and the depressurization valves (DPVs) should be inevitable to lead to large LOCA even for the case of small LOCA by forcibly opening the DPVs to inject water from the GDCS pool due to that the GDCS water head is up to ∼0.2MPa. On the contrary, our simulation exhibited that SI-PCIS could save the reactors from leading to large LOCA by discharge of the water into a core for the cases of small LOCA or DPV unexpectedly open. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.

Experiments and Analytical Simulation Work on an Innovative Steam-Injector-Driven Passive Core Injection Cooling System

2007 ◽  
Author(s):  
Michitsugu Mori ◽  
Tadashi Narabayashi ◽  
Shuichi Ohmori ◽  
Fumitoshi Watanabe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Pressures ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Coolant Injection ◽  
Core Cooling

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

Hybrid heat pipe based passive in-core cooling system for advanced nuclear power plant

2015 ◽  
Vol 90 ◽  
pp. 609-618 ◽  
Author(s):  
Yeong Shin Jeong ◽  
Kyung Mo Kim ◽  
In Guk Kim ◽  
In Cheol Bang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Pipe ◽  
Nuclear Power ◽  
Cooling System ◽  
Core Cooling

Optimization of Test Interval of Ignalina Nuclear Power Plant Auxiliary Feedwater Pumps

2009 ◽  
Author(s):  
Roman Voronov ◽  
Robertas Alzbutas
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Failure Modes ◽  
Cooling System ◽  
Model Parameters ◽  
Testing Interval ◽  
Safety Function ◽  
Core Cooling ◽  
Safety Systems

Some safety systems of the Ignalina Nuclear Power Plant (NPP) operate in standby mode. An equipment of such systems is periodically tested and that allows timely detect and repair equipment failures. The periodic testing is an important measure of ensuring systems’ operability and reliability. However, during the test and repair the equipment cannot perform it’s safety function, therefore too often testing decreases the availability of the system. This paper describes the mathematical model that represents how availability and reliability of the systems and their components depend on testing interval, taking into account different failure modes of the equipment. This model allows to find the optimal testing interval for the safety. As an example, the auxiliary feedwater pumps, that are a part of the Ignalina NPP Reactor Emergency Core Cooling System, are analysed. The model parameters calculation is based on Ignalina NPP data regarding pumps operation and failure as well as on general Nordic NPPs reliability data (T-Book) appling Bayesian approach for parameters updating. The analysed safety system is a redundant system that consists of six pumps and other equipment. Therefore a model for multiple components failure was developed. The model accounts for actual operational requirements of the system. The results of this model are compared with usually used binominal model.

Assessment of Reliability of Emergency Core Cooling System (ECCS) of Bushehr Nuclear Power Plant

2009 ◽  
Author(s):  
Mohammad Sotoudeh ◽  
Kamran Sepanloo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Computer Code ◽  
Probabilistic Methods ◽  
Event Tree ◽  
Loss Of Coolant Accident ◽  
Fuel Storage ◽  
Core Cooling

Bushehr Nuclear Power Plant (BNPP) originally designed by German KWU and then modified by Russian companies is approaching commissioning. In this paper, the reliability of Emergency Core Cooling System (ECCS) of the new design is compared with the older one using both deterministic and probabilistic methods. The reliability of both systems is calculated in case of occurrence of Large Loss of Coolant Accident (Large LOCA). The new Russian design is based on the original KWU Convoy design and the applied modifications have improved its reliability. To perform the reliability analysis the event tree/fault tree method is used and the calculations are done by IAEA computer code PSAPACK. The results show that the applied modifications such as, increase of redundancy from 4 × 50% to 4 × 100%, change of cool down route, fuel storage pool cooling in all four trains have reduced the ECCS unavailability from 5.9 E−4 to 1.7 E−4. Furthermore, based on the results it is shown that both designs comply well with the IAEA recommendations on Probabilistic Safety Criteria (PSC) and INSAG-3 requirements.

Development and D3 Application of ABWR Feedwater System Model

2009 ◽  
Author(s):  
Hui-Wen Huang ◽  
Chunkuan Shih ◽  
Hung-Chih Hung ◽  
Kai-Lan Chang ◽  
Shu-Chuan Chen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Simulation Code ◽  
Actuation System ◽  
Feedwater Pump ◽  
Depth Analysis ◽  
Safety Features ◽  
Core Cooling

This work developed an Advanced Boiling Water Reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element / three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including inadvertent closure of all turbine control valves and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an Emergency Core Cooling System (ECCS) automatic diverse back-up. In Lungmen Nuclear Power Plant (NPP), a diverse manual initiation means for the High Pressure Core Flooder (HPCF) loop C is designed as the back-up of digitalized Engineered Safety Features Actuation System (ESFAS). If the Motor-Driven Feedwater Pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common cause failure (CCF).

The Effects of Subcooling on Quenching of a Vertical Brass Cylinder With Heating Power

2015 ◽  
Author(s):  
Yuan-Hong Ho ◽  
Ming-Xi Ho ◽  
Chin Pan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Fission Products ◽  
Heating Power ◽  
Film Boiling ◽  
Cylinder Surface ◽  
Constant Wall Temperature ◽  
Core Cooling

Film boiling is usually induced while a very hot object contacts with a coolant. Such phenomena will deteriorate the heat transfer and degrade the cooling process. Film boiling is of significant concern for the design of an emergency core cooling system after a hypothetical loss of coolant accident happens in a nuclear power plant. Furthermore, after a nuclear power plant is shut down, the fuel rods will continue to release heat due to the decay of fission products. Moreover, the subcooling of coolant might be changed dramatically during the reflood process. Therefore, it is of significant importance and interest to understand the effect of decay heat and subcooling of coolant on the quenching process of a hot object. This study demonstrates the quenching of a vertical brass cylinder without and with heating power of 105W in deionized water with different subcoolings. The diameter and length of the cylinder is 24 mm and 112 mm, respectively. Six K-Type thermocouples are embedded 2mm below the cylinder surface at different axial locations. The experimental results reveal that, with heating power of 105W, the duration of film boiling becomes much larger than the case without heating power under the same subcooling condition. Besides, the duration of film boiling decreases with increasing subcooling. This study also reveals that the Leidenfrost temperature increases significantly with increasing the subcooling with or without heating power. Significantly, a stable film boiling with approximately constant wall temperature can be sustained in saturated water. This is of significant concern for nuclear safety.

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