Study on Operation Ranges and Discharge Performance of Steam Injector System

2017 ◽  
Author(s):  
Takahiro Moribe ◽  
Hiroto Endo ◽  
Shuichiro Miwa ◽  
Hiroto Sakashita ◽  
Michitsugu Mori
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Flow ◽  
Flow Rate ◽  
Nuclear Power ◽  
Cooling System ◽  
Water Flow Rate ◽  
Steam Pressure ◽  
Pump Performance ◽  
Operating Range

Steam Injector (SI) is a passive pump activated by steam and water, and it does not require any external power supplies or rotating machineries. Moreover, SI has high ability as a heat exchanger by undergoing direct contact condensation mechanism. From these characteristics, SI has a capability to be applied as a passive safety system in the nuclear power plant. In the present study, experiments targeting the operating range and pump performance were carried out to obtain SI’s detailed characteristics under following supply conditions; inlet steam pressure was 0.02 ∼ 0.81MPaG, inlet water flow rate was 0.21 ∼ 0.80 kg/s. In the former experiment, operating range was investigated by changing inlet conditions, and the influence of steam inlet pressure and water inlet flow rates on SI operating range was tested. In the latter experiment, pump performance of SI was evaluated by investigating the maximum discharge pressure in each inlet condition by using back pressure valve. From the results of experiments, it was confirmed the operating range of SI was limited by supplied steam pressure and supplied water flow rate, and some clear trends were found in the operation boundary map. In addition, SI could discharge water at least 1.2 times higher than inlet steam pressure under above-mentioned conditions, which verified SI’s capability to be applied for the nuclear power plant as a core cooling system. And furthermore, existing 1D analysis model’s predictive capability was tested based on these experimental results.

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.

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Vanderley Vasconcelos ◽  
Wellington Antonio Soares ◽  
Raissa Oliveira Marques ◽  
Silvério Ferreira Silva Jr ◽  
Amanda Laureano Raso
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Cooling System ◽  
Human Reliability ◽  
Non Destructive ◽  
Plant Components

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

Bratislavia dadayi (Michaelsen 1905) (Annelida, Clitellata, Naididae): discovery of an alien oligochaete in a technogenic fresh water body in Ukraine

Zootaxa ◽  
2019 ◽  
Vol 4711 (2) ◽  
pp. 349-365
Author(s):  
VLADIMIR A. GUSAKOV ◽  
ANZHELIKA A. SYLAIEVA
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fresh Water ◽  
Water Body ◽  
Nuclear Power ◽  
Potential Distribution ◽  
Cooling System ◽  
European Continent ◽  
Previous Record ◽  
Sexually Mature

A non-native oligochaete, Bratislavia dadayi (Michaelsen 1905), is recorded from a water body of the cooling system of the Khmelnitsky Nuclear Power Plant (Ukraine). This is the first registration of this species in the central part of the European continent, far from sea and river navigable waterways. The only previous record of B. dadayi in Europe had been from a Belgian estuary. The occurrence in samples taken over several years, and the presence of sexually mature individuals in the Ukrainian population indicate the worm’s successful naturalization in the new habitat. In this paper, we analyze the species’ morphology and abundance in the Ukrainian population and discuss its ecology, current and potential distribution. 

Dynamic Buckling of Axisymmetric Shells

1974 ◽  
Vol 41 (4) ◽  
pp. 1063-1068 ◽  
Author(s):  
A. Kalnins
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Free Vibration ◽  
Dynamic Stability ◽  
Nuclear Power ◽  
Cooling System ◽  
Dynamic Buckling ◽  
Reference Surface ◽  
Long Duration ◽  
Axisymmetric Shells

A procedure for the analysis of dynamic buckling of axisymmetric shells subjected to axisymmetric, periodic loads of long duration is proposed that is based on the calculation of the nonsymmetric modes of free vibration and associated mode integrals over the reference surface of the shell. Numerical results are presented for the evaluation of dynamic stability of an actual shell that is designed for the cooling system of a nuclear power plant.

Estimation of Optimal Water Flow Rate of a Steel Roller Shutter with Water Film Cooling System

2014 ◽  
Vol 905 ◽  
pp. 263-267
Author(s):  
Shin Ku Lee ◽  
W.H. Lo ◽  
M.C. Ho ◽  
T.H. Lin
Keyword(s):  
Water Flow ◽  
Flow Rate ◽  
Film Cooling ◽  
Inverse Method ◽  
Cooling System ◽  
Water Flow Rate ◽  
Water Film ◽  
Steel Roller ◽  
Fire Environment ◽  
Hot Surface

The hybrid inverse method to estimate the optimal water flow rate and surface temperature on the hot surface of the steel roller shutter with water film cooling system subjected to a fire environment is presented in this paper. The results show that the effect of the down-flowing water film flow rate on the present estimates cannot be negligible. The water-film system combined with the steel roller shutter can effectively improve the heat resistance and the temperature of the shutter slat surface can be controlled to around 100 °C. The optimal water flow rate is 110 L/min for a typical 3m x 3m steel roller shutter with water film cooling system.

Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit

2020 ◽  
Author(s):  
Shengtao Zhang ◽  
Ke Yi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Cooling System ◽  
Water System ◽  
Total Loss ◽  
Operating Strategy ◽  
Service Water ◽  
Essential Service

Abstract Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient. This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in Generation III nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.

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