Study on Unidirectional Solidification of Slab Ingot for Special Heavy Plate

2011 ◽  
Vol 189-193 ◽  
pp. 4033-4036
Author(s):  
Xing Dong Peng ◽  
Lin Hu ◽  
Pei Xin Wang ◽  
Lian Gang Zhao
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Simulation Experiment ◽  
Unidirectional Solidification ◽  
Hydropower Station ◽  
Aircraft Carrier ◽  
Oil Drilling ◽  
Heavy Plate ◽  
Large Size ◽  
Pass Through

Unidirectional solidification is a method of producing slab ingot for special heavy plate, when compression radio is more than 2.5, internal soundness of slab ingot is assurance, and the nature of Z orientation and percent of pass through supersonic flaw detecting are increased. Products are used in oil drilling platform in ocean ,separator of steam turbo-alternator with more than 0.6 million kW, anchor gate of large-scale hydropower station, shell of nuclear power plant, armoured plate of aircraft carrier. According to dissection on 28t large-sized slab ingot, simulation experiment of crystal, and analogy calculation of heat transfer and coagulation with CFX software, histology features and key to manufacturing technology of large-size unidirectional solidification slab ingot were mastered.

New Construction Technology for Condenser Shell Module of Large Scale Nuclear Power Station

2012 ◽  
Vol 268-270 ◽  
pp. 1039-1043
Author(s):  
Yong Jun Xia ◽  
Yan Long Wang ◽  
Yan Hui Yin ◽  
Qian Miao
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Module Structure ◽  
Construction Technology ◽  
Large Size ◽  
Construction Scheme ◽  
New Construction ◽  
Traditional Construction ◽  
House Structure ◽  
Power Project

The shell is one important part of steam turbo-set condenser. Fangjiashan nuclear power project in Zhejiang Province is the domestic millions-kilowatt class PWR nuclear power plant. Its condenser shell modules have large size and heavy weight. At the same time by turbine house structure restriction, the traditional construction scheme is unable to satisfy the installation requirements. According to the condenser shell module structure characteristics and construction environment, the new construction technology such as four cranes integration operation, many times of changing hook, temporary hanging on the beam of steam turbine house, is studied and adopted. The supporting lifting frame is researched and developed. Four pieces of condenser shell modules with different hanging point size are successfully constructed. This can provide reference for the construction of similar equipment.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

Agricultural Land Management Options Following Large-Scale Environmental Contamination: Evaluation for Fukushima Affected Agricultural Land

2013 ◽  
Author(s):  
Hildegarde Vandenhove
Keyword(s):  
Land Use ◽  
Land Management ◽  
Nuclear Power ◽  
Large Scale ◽  
Agricultural Land ◽  
Agricultural Practices ◽  
Management Option ◽  
Management Options ◽  
Fukushima Daiichi ◽  
Agricultural Land Management

The accident at the Fukushima Daiichi Nuclear Power Plant has raised questions about the accumulation of radionuclides in soils, the transfer in the foodchain and the possibility of continued restricted future land use. This paper summarizes what is generally understood about the application of agricultural countermeasures as a land management option to reduce the radionuclides transfer in the food chain and to facilitate the return of potentially affected soils to agricultural practices in areas impacted by a nuclear accident.

Parts of South-east Asia may adopt nuclear post-2030

2021 ◽  
Keyword(s):  
East Asia ◽  
Nuclear Power ◽  
Large Scale ◽  
East Asian ◽  
Energy Development ◽  
South East Asia ◽  
Asian Countries ◽  
Content Type ◽  
Electricity Grid ◽  
Renewable Energy Development

Significance It is the only country in South-east Asia with a large-scale nuclear plant, although this was never loaded with fuel. Other countries in the region have tentative plans to develop nuclear power programmes. Impacts The current absence of nuclear power programmes will help avert the diversion of capital from renewable energy development in the region. South-east Asian countries with small, non-power reactors, built for research, will try to maintain these facilities. Across the region, the need for electricity grid investment will increase as more decentralised generation sources are deployed.

Development of 1500-r/min 75-Inch Ultra-Long Last Stage Blades for Nuclear Steam Turbine

2017 ◽  
Author(s):  
Deqi Yu ◽  
Jiandao Yang ◽  
Wei Lu ◽  
Daiwei Zhou ◽  
Kai Cheng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Vibration Monitoring ◽  
Steam Turbines ◽  
Element Analysis ◽  
Rotating Blade ◽  
Lifetime Analysis ◽  
Computational Fluid Dynamics Cfd ◽  
Last Stage

The 1500-r/min 1905mm (75inch) ultra-long last three stage blades for half-speed large-scale nuclear steam turbines of 3rd generation nuclear power plants have been developed with the application of new design features and Computer-Aided-Engineering (CAE) technologies. The last stage rotating blade was designed with an integral shroud, snubber and fir-tree root. During operation, the adjacent blades are continuously coupled by the centrifugal force. It is designed that the adjacent shrouds and snubbers of each blade can provide additional structural damping to minimize the dynamic stress of the blade. In order to meet the blade development requirements, the quasi-3D aerodynamic method was used to obtain the preliminary flow path design for the last three stages in LP (Low-pressure) casing and the airfoil of last stage rotating blade was optimized as well to minimize its centrifugal stress. The latest CAE technologies and approaches of Computational Fluid Dynamics (CFD), Finite Element Analysis (FEA) and Fatigue Lifetime Analysis (FLA) were applied to analyze and optimize the aerodynamic performance and reliability behavior of the blade structure. The blade was well tuned to avoid any possible excitation and resonant vibration. The blades and test rotor have been manufactured and the rotating vibration test with the vibration monitoring had been carried out in the verification tests.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

2008 ◽  
Author(s):  
Mitsuhiro Suzuki ◽  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Simulation Experiment ◽  
Three Dimensional ◽  
Test Facility ◽  
Large Temperature ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Accident Management

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

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