Effect of Opening on the Response of Containment Model to Pressure Loading

2006 ◽  
Author(s):  
M. K. Agrawal ◽  
A. Ravi Kiran ◽  
A. K. Ghosh ◽  
H. S. Kushwaha
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Dead Weight ◽  
Modeling Methodology ◽  
Containment Model ◽  
Scale Ratio ◽  
Heavy Water Reactor ◽  
Set Up ◽  
Accident Conditions ◽  
Design Pressure

The Containment Studies Facility (CSF) is being set up in BARC for studying various containment related thermal hydraulic and other phenomena which occur during simulated accident conditions in Nuclear power Plants. The facility consists of a concrete containment model having a volumetric scale ratio of 200:1 with respect to the actual containment of Indian Pressurized Heavy Water Reactor. The structure is designed for pressure of 1.73 Kg/cm2 for specified leak tightness. Adequacy to withstand design pressure is checked by test as well as numerical analysis before commissioning of the facility. Accordingly Containment building model has been analyzed by finite element method for internal design pressure and dead weight. Analysis has been carried out for the structure with and without the opening in the containment. Effect of opening on the response of containment has been studied. The paper includes the modeling methodology, maximum deflection and stress amplification around the opening for various models.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

scholarly journals Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

2020 ◽  
Author(s):  
Evrim Oyguc ◽  
Abdul Hayır ◽  
Resat Oyguc
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Energy Sources ◽  
Finite Element Program ◽  
Design Spectrum ◽  
Reactor Building ◽  
Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

Experimental Study on the Interaction of Molten Sn With Water

2018 ◽  
Author(s):  
Longkun He ◽  
Pengfei Liu ◽  
Xisi Zhang ◽  
Wenjun Hu ◽  
Bo Kuang ◽  
...  
Keyword(s):  
Water Temperature ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Dynamic Pressure ◽  
Ambient Pressure ◽  
Influence Factors ◽  
Video Camera ◽  
Metallic Oxide ◽  
Set Up

In nuclear power plants, fuel-coolant interaction (FCI) often accompanied with core melt accidents, which may escalate to steam explosion destroying the integrity of structural components and even the containment under certain conditions. In the present study, a new facility for intermediate-scaled experiments named ‘Test for Interaction of MELt with Coolant’ (TIMELCO) has been set up to study FCI phenomena and thermal-hydraulic influence factors in metal or metallic oxide/water mixtures with melt at maximum 2750°C. The first series of tests was performed using 3kg of Sn which was heated to 800°Cand jetted into a column of 1m water depth (300mm in diameter) under 0.1MPa ambient pressure. The main changing parameter was water temperature, at 60 °C and 72 °C respectively. From the high-speed video camera, violent explosion phenomenon occurred at water temperature of 60°C, while no evident explosion observed at 72°C. The size of melt debris at 60°C is smaller than this at 72°C.On the contrary, the dynamic pressure at 60°C is larger. The results indicate that water temperature has an important effect on FCI and decreasing the temperature of the coolant is advantageous to the explosion.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

Research on Combined Peak Load Regulation with Hydropower, Thermal Power and Nuclear Power Plants

2014 ◽  
Vol 986-987 ◽  
pp. 465-469 ◽  
Author(s):  
Gang Wang ◽  
Xiao Dong Ma ◽  
Chao Wang ◽  
Peng Ye
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Power Grid ◽  
Thermal Power ◽  
Operation Mode ◽  
Peak Load ◽  
Generating Capacity ◽  
Load Regulation ◽  
Set Up ◽  
Research Show

Base on Nuclear Power Plant (NPP) participating in peak load regulation of power grid, this paper studies the operation mode of hydropower, thermal power and NPP in Combined Peak Load Regulation. The optimization model for Peaking depth of NPP was set up. The case based on actual power grid were calculated and analyzed, results of the research show that in combined peak load regulation of hydropower, thermal power and NPP, a reasonable peaking depth of NPP will effectively alleviate the peaking pressure of power grid, avoid start-stop of thermal power and abandoned water of hydropower, while ensuring the hydroelectric generating capacity in the low load periods, and ensure thermal power output smooth, it further reduce the operating costs, verify the effectiveness of the model.

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