Detailed Experimental and Analytical Study on Long Two-Phase Closed Thermosiphons Related to Passive Spent-Fuel Pool Cooling

2018 ◽  
Author(s):  
Claudia Graß ◽  
Anne Krüssenberg ◽  
Rudi Kulenovic ◽  
Fabian Weyermann ◽  
Jörg Starflinger ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
High Heat ◽  
Heat Transfer Process ◽  
Normal Operation ◽  
Working Fluid ◽  
Spent Fuel ◽  
Two Phase ◽  
Spent Fuel Pool ◽  
Residual Heat

New concepts are currently being discussed for passive residual-heat removal with heat pipes from spent-fuel pools and wet-storage facilities. Because of their high heat-transport capability and their simple design, two-phase closed thermosiphons have a great potential to satisfy the demands of a reliable and independent passive heat removal. The geometry of spent-fuel pools and the potential incorporation into existing plants requires thermosiphons of at least 10 m in length including bends. Such thermosiphons are neither available nor have they been investigated yet. Therefore, experimental and numerical investigations are being carried out. At IKE the basic operational behavior of 10-m-long thermosiphons with water — as working fluid — are being experimentally investigated. Measurements for different pipe diameters (32 mm and 45 mm) are performed at various heat sink temperatures (10 °C, 20 °C and 30 °C), heat inputs (1000 W to 4000 W), and filling ratios (50%, 70% and 100%). GRS is developing codes, such as AC2, in order to simulate all relevant phenomena within a nuclear power plant during normal operation, incidents, accidents, and severe accidents. Regarding passive residual-heat removal with thermosiphons, the models of AC2 are being improved to properly simulate the thermos-hydraulics of this heat transfer process. Starting with the module ATHLET (Analysis of Thermal-hydraulic of Leaks and Transients), the applicability of its existing models is checked for modeling long thermosiphons and calculating their operational behavior. The main model improvements are being validated against the new experiments of IKE.

Study on Water Hammer in the Overpressure Discharge Pipe of Nuclear Power Plant

2017 ◽  
Author(s):  
Qingmu Xu ◽  
Kun Cai ◽  
Jie Qin ◽  
Junkai Yuan ◽  
Juan Li
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Hammer ◽  
Liquid Column ◽  
Normal Operation ◽  
Pipeline System ◽  
Spent Fuel ◽  
Vapor Cavity ◽  
Spent Fuel Pool

Water hammer phenomenon is a significant pressure wave in pipe system caused by momentum change when the moving fluid is forced to stop or change direction instantaneously. Common causes of water hammer are sudden valve closing at the end of a pipeline system, pump failure, check valve slam etc. The steam transportation pipeline system may also be vulnerable to water hammer when it confronts with the situation where liquid and steam co-exist. Water hammer often occurs when steam condenses into water in a horizontal section of steam piping. Then steam “picks up” water to form a high-velocity “slug” and create extra stress to pipe. When steam is trapped into sub-cooled water, the collapse of vapor cavity can lead to collision of two columns of liquid, resulting in a large rise in pressure which will damage pipes, supporting structures and hydraulic machinery. Nuclear power plant is composed of complex equipments and piping systems, lots of which contain both liquid and steam. Hence, there is a potential threat of occurrence of water hammer to the normal operation of systems. Thus, this phenomenon needs to be well investigated and prevented with some effective methods. For the purpose of overpressure relief under severe accidents, the spent fuel pool cooling system of CAP1000 series nuclear power plant provides a discharge passage from containment to spent fuel pool. When the containment pressure exceeds the control value, valve is opened to discharge high-temperature and high-pressure steam until the pressure drops to a safety value. During this process, serious water hammer happens, causing pressure rise beyond the design pressure and further leading to damages to pipes and structures. Therefore, water hammer of overpressure discharge pipeline in CAP1000 plant is studied in this work. On the basis of verification of the capabilities of computational code RELAP5/MOD3.3, hydraulic transient of water hammer is simulated under different conditions. It is indicated that after steam discharge stops, residual steam in pipe condenses because of contact with sub-cooled water in spent fuel pool. Subsequently, the rapid backflow and vapor cavity lead to a severe water hammer. The detailed analysis has shown that water temperature of spent fuel pool has a decisive influence on the mechanism of water hammer phenomenon, including collision of liquid column to valve disc and cavity collapse in the horizontal pipe. The collision and separation of liquid column result in relatively lower pressure amplitude.

Nuclear Power Plant Passive Residual Heat Removal System Design and Simulation

2013 ◽  
Author(s):  
Xiao Yuan ◽  
Minjun Peng ◽  
Genglei Xia
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Primary Circuit ◽  
Two Phase ◽  
Pressurized Water ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

The passive safety systems employed in the design of pressurized water reactor (PWR) can accomplish the inherent safety functions and mitigate the consequences of the postulated accidents. In this paper, a passive residual heat removal system (PRHRs) is designed for a certain nuclear power plant. The RELAP5/MOD3.4 code was used to analyze the operation characteristics of the PRHRs. It shows the PRHRs could remove the decay heat from the primary loop effectively, and the single-phase and two-phase natural circulations could respectively establish in the primary circuit and the PRHRs circuit.

scholarly journals Design, Experiment, and Commissioning of the Passive Residual Heat Removal System of China’s Generation III Nuclear Power HPR1000

2021 ◽  
Vol 2021 ◽  
pp. 1-6
Author(s):  
Feng Li ◽  
Yazhe Lu ◽  
Xiao Chu ◽  
Qiang Zheng ◽  
Guanghao Wu
Keyword(s):  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Two Phase ◽  
Safe State ◽  
Station Blackout ◽  
Run Test ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

In response to a station blackout accident similar to the Fukushima nuclear accident, China’s Generation III nuclear power HPR1000 designed and developed a passive residual heat removal system connected to the secondary side of the steam generator. Based on the two-phase natural circulation principle, the system is designed to bring out long-term core residual heat after an accident to ensure that the reactor is in a safe state. The steady-state characteristic test and transient start and run test of the PRS were carried out on the integrated experiment bench named ESPRIT. The experiment results show that the PRS can establish natural circulation and discharge residual heat of the first loop. China’s Fuqing no. 5 nuclear power plant completed the installation of the PRS in September 2019 and carried out commissioning work in October. This debugging is the first real-world debugging of the new design. This paper introduces the design process of the PRS debugging scheme.

An Experimental Study of Closed Loop Two-Phase Thermosyphon for Spent Fuel Pool Passive Cooling

2016 ◽  
Author(s):  
Minglu Wang ◽  
Mingguang Zheng ◽  
Cheng Ye ◽  
Zhongming Qiu ◽  
Zhenqin Xiong
Keyword(s):  
Heat Transfer ◽  
Experimental Study ◽  
Heat Transfer Coefficient ◽  
Thermal Resistance ◽  
Transfer Coefficient ◽  
Closed Loop ◽  
Working Fluid ◽  
Spent Fuel ◽  
Two Phase ◽  
Spent Fuel Pool

The study reported here examined a closed loop two-phase thermosyphon (CLTPT) of evaporator length 7.6m and internal diameter 65mm used to cool the spent fuel pool. This experimental study investigates the thermal performances and heat and mass transfer characteristic of CLTPT by examining the thermodynamic cycles and overall thermal resistances with ammonia, R134a and water as the working fluid. Measurements of temperature and pressure distributions of the fluid around the loop were made under various conditions. Results show that this loop operates with low filling ratio, low mass flow rate, and high heat-transfer coefficient and the CLTPT has the ability to cool the spent fuel pool. The working fluid flowing through the heat pipe evaporator section generally experienced a subcooled zone, pool boiling zone and high gas quality two-phase region. The average heat transfer coefficient of evaporator reaches 450 W/m2•°C using R134a as working fluid. The thermal resistance of R134a is always smaller than ammonia but the thermal resistance of water is largest at small temperature difference while is smallest when temperature difference is large.

Performance Analysis of a Novel PXS With Organic Rankine Cycle (ORC) System

2013 ◽  
Author(s):  
Shengjun Zhang
Keyword(s):  
Nuclear Power ◽  
Organic Rankine Cycle ◽  
Heat Removal ◽  
Rankine Cycle ◽  
Working Fluid ◽  
Heat Transfer Area ◽  
Decay Heat ◽  
The Core ◽  
Primary Fluid ◽  
Residual Heat

With the increasing of core thermal power of the nuclear power plant, the decay heat of the core increases in the accident. Therefore, the heat removal capacity of the PXS should be enhanced to fulfill the requirement of core safety. A new scheme is put forward to improve the cooling capacity of PXS and provides long-term power for station blackout (SBO) accident or loss of normal feedwater. In this system, the Organic Rankine Cycle is incorporated between the hot leg and cold leg of PRHR. The decay heat of the core is the heat source and the cooling pool outside the containment is the cool source. The natural circulation of the primary loop is established due to the density difference. The primary fluid flows into the evaporator of the ORC system, where the working fluid of the ORC system is evaporated. Then the temperature of the primary fluid is decreased. The vaporized working fluid drives the expander, which is coaxially fixed with the fluid pump, to generate the power. Finally, the exhausted vapor flows into the condenser and the residual heat is discharged outside of the containment simultaneously. The working fluid in the condenser is pumped into the evaporator by the fluid pump for liquid supplement and the cycle keeps on working continuously. A steady state analysis is performed on a 1700MWe nuclear power plant with ORC as the heat removal system. The heat transfer area of the ORC evaporator is fixed as 487.7m2, which is the same as the area of PRHR HX. The efficiencies of fluid pump and expander of ORC system are assumed as 0.75 and 0.8, respectively. The decay heat of the core is about 67.62MWe, which is 1.38% of the core full power. The working fluids are screened and R141b offers excellent performance. The efficiency of fluid pump and expander are assumed as 0.75 and 0.8, respectively. The condensing temperature is assumed as 80°C and the evaporating temperature is 160°C. The results show that 7.83MWe will be generated by the ORC system and the heat transfer area of the condenser is about 994.5m2. The residual heat of 59.79MWe will be discharged to the water tank outside the containment.

The holistic risk analysis of spent fuel pool considering the fuel route risk

2021 ◽  
pp. 1748006X2199851
Author(s):  
Zhixin Xu ◽  
Ming Wang ◽  
Binyan Song ◽  
WenYu Hou ◽  
Chao Wang
Keyword(s):  
Nuclear Power ◽  
Comprehensive Evaluation ◽  
Point Of View ◽  
Spent Fuel ◽  
Nuclear Disaster ◽  
Evaluation Approach ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Large Water ◽  
Fukushima Nuclear Disaster

The Fukushima nuclear disaster has raised the importance on the reliability and risk research of the spent fuel pool (SFP), including the risk of internal events, fire, external hazards and so on. From a safety point of view, the low decay heat of the spent fuel assemblies and large water inventory in the SFP has made the accident progress goes very slow, but a large number of fuel assemblies are stored inside the spent fuel pool and without containment above the SFP building, it still has an unignored risk to the safety of the nuclear power plant. In this paper, a standardized approach for performing a holistic and comprehensive evaluation approach of the SFP risk based on the probabilistic safety analysis (PSA) method has been developed, including the Level 1 SFP PSA and Level 2 SFP PSA and external hazard PSA. The research scope of SFP PSA covers internal events, internal flooding, internal fires, external hazards and new risk source-fuel route risk is also included. The research will provide the risk insight of Spent Fuel Pool operation, and can help to make recommendation for the prevention and mitigation of SFP accidents which will be applicable for the SFP configuration risk management.

scholarly journals Microbial enrichment and gene functional categories revealed on the walls of a spent fuel pool of a nuclear power plant

PLoS ONE ◽  
2018 ◽  
Vol 13 (10) ◽  
pp. e0205228 ◽  
Author(s):  
Rosane Silva ◽  
Darcy Muniz de Almeida ◽  
Bianca Catarina Azeredo Cabral ◽  
Victor Hugo Giordano Dias ◽  
Isadora Cristina de Toledo e Mello ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Spent Fuel ◽  
Functional Categories ◽  
Spent Fuel Pool
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