scholarly journals Analysis on flow induced vibration of main steam bypass valve in a nuclear power plant

2021 ◽  
Vol 1985 (1) ◽  
pp. 012027
Author(s):  
Junfeng Liu ◽  
Xudong Chen ◽  
Chuangao Han ◽  
Xiaolong Ma ◽  
Fangjie Wu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Flow Induced Vibration ◽  
Bypass Valve ◽  
Main Steam

Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Zhou Gengyu ◽  
Liang Shuhua ◽  
Sun Lin ◽  
Lv Feng
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Load Capacity ◽  
Operating Conditions ◽  
Integrity Assessment ◽  
Inner Radius ◽  
Main Steam

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users. This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.

Investigation of a Possible Emergency Procedure for the VVER 1000 NPP in Case of a Total Loss of Feedwater and a Main Steam Line Break

2004 ◽  
Author(s):  
Nikolaus Muellner ◽  
Walter Giannotti ◽  
Francesco D’Auria
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Total Loss ◽  
Normal Operation ◽  
Feed Water ◽  
Emergency Procedures ◽  
Design Basis ◽  
Management Procedures ◽  
Main Steam

Accident management procedures associated with nuclear power plant beyond design basis accidents should be developed with the aid of the more recent versions of advanced computational tools (best estimate codes) in order to verify the effectiveness of normal operation systems of the plant to avoid or minimize core damage. A. Madeira showed in her paper “A PWR Recovery Option for a Total Loss of Feedwater Beyond Design Basis Scenario” that it is possible to safely control a total loss of feedwater scenario in the Angra2 nuclear power plant, using two emergency procedures, namely the opening of the steam generator (SG) relief valves, and on the primary side the complete manual opening of all pressurizer relief and safety valves. This paper investigates the effectiveness of the procedure opening of the SG relief valves, followed by primary side feed and bleed for a generic VVER-1000 NPP in case of a total loss of feed water. The results indicate that the procedure is successful in reducing the primary side pressure and temperature to safe conditions, i.e. long term core cooling is achievable.

Simulation of the Acoustic Loads Generated in the Intersection of a Main Steam Line and its Safety Relief Valve Branch of a BWR Plant under Extended Power Uprate Condition

2015 ◽  
Vol 362 ◽  
pp. 190-199
Author(s):  
Arturo Ocampo-Ramirez ◽  
Luis Héctor Hernández-Gómez ◽  
Pablo Ruiz-López ◽  
Noel Moreno-Cuahquentzi ◽  
Guillermo Urriolagoitia-Calderón ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
The United States ◽  
Steam Flow ◽  
Normal Operation ◽  
Relief Valve ◽  
Steady State Condition ◽  
Main Steam ◽  
Extended Power

The structural integrity of a BWR nuclear power plant can be compromised due to severe dynamic loads. Acoustic loads coming from a Safety Relief Valve branch can be adversely amplified if the steam flow is increased at the Main Steam Piping. This phenomenon has been reported previously in one BWR nuclear power plant. Its steam dryer was fractured and loose parts were generated due to high-cycle fatigue. This event has driven the United States Nuclear Regulatory Commission to issue specific regulations to evaluate acoustic loads which would be detrimental to the BWR steam dryer. In this paper, the acoustic loads were simulated when the steam flow is incremented from normal operation conditions to an Extended Power Uprate condition. It was analyzed, when the output power was incremented 14% and 28%. The initial conditions were determined with Computational Fluid Dynamics under steady state condition. This data was used in subsequent transient analysis. The model of Large Eddy Simulation was used and the acoustic simulation was performed with the Fowcs Williams and Hawkings Method. The Power Spectral Density was obtained with Fast Fourier Transform. The frequency peaks were found between 148 Hz and 155 Hz. These results are consistent with those obtained with the Helmholtz model and other results reported in the open literature. The results show that the peak pressure can be increased up to six times in resonance conditions, corresponding to a power uprate of 28%.

Thermal Hydraulic Performance Analysis for Passive Containment Cooling System of AP1000 in the Event of LOCA

2017 ◽  
Author(s):  
Talha Bin Mujahid ◽  
Yu Yu ◽  
Bin Wang ◽  
Muhammad Ali Shahzad ◽  
Fenglei Niu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Analysis ◽  
Hydraulic Performance ◽  
Steam Line ◽  
Loss Of Coolant ◽  
Main Steam Line ◽  
Main Steam

The design of a nuclear reactor containment building is of key importance in order to enhance the safety of a nuclear power plant. Owing to nuclear accidents such as TMI, Chernobyl and Fukushima, more and more attention is paid to the passive concept in nuclear power development. In order to improve the safety of new generation nuclear power plant, passive systems are widely used, passive containment cooling system in AP1000 is one of the typical example of such kinds of systems. It’s function is to transfer the heat produced in the containment to the atmosphere and keep the pressure in the vessel below the threshold under such accidents as Loss of coolant (LOCA), main steam line break (MSLB), etc. The system operates based on natural circulations inside the steel vessel and in the air baffle outside the containment, and the cooling water is sprayed to the steel surface to enhance the heat transfer process. A proper model simulating the system behavior is needed for system design and safety analysis, and a multivolume lumped parameter approach is employed in order to analyze the containment integrity and to study the long term response of postulated Loss of coolant (LOCA) accidents and Main steam line break (MSLB) accidents. However, the temperature and pressure distributions cannot be described detailed by such model, which is important to study the T-H characteristics in the containment. In this paper LOCA has been simulated on MATLAB using a given pipe break size and the response of containment is analyzed. Furthermore, the results are compared with the results in the Westinghouse Design Control Document 2002. Then the thermal hydraulic performance is studied, the factors such as the air temperature, containment pressure and mass flow rate of the coolant and their effects on the containment are analyzed. This research is done to get further insight on the safety analysis of reactor containment regarding maximum temperature and stress calculation inside the containment.

EXPERIMENTAL STUDY ON IMPACT FRETTING WEAR OF INCONEL TUBES UNDER HIGH TEMPERATURE AND PRESSURE

2011 ◽  
Vol 25 (31) ◽  
pp. 4253-4256 ◽  
Author(s):  
CHOON YEOL LEE ◽  
JOON WOO BAE ◽  
YOUNG SUCK CHAI ◽  
KYOOSIK SHIN
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Impact Force ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Experimental Parameters ◽  
Generator Design ◽  
The Impact

In nuclear power plant, fretting wear caused by flow induced vibration (FIV) accompanied with impact force can make serious problems between U -tubes and egg-crates which are located in steam generators. In order to guarantee the reliability of the steam generator, design based on consideration of the damage due to the fretting wear of the U -tube is inevitable. The purpose of this study is to elucidate fretting wear mechanism qualitatively and quantitatively. First, finite element models are developed to analyze the dynamic characteristics and estimate the impact force in steam generators. Based on the numerical results, fretting wear simulation is performed according to the environment to which the actual steam generators in nuclear power plant are exposed. Initial experimental results are obtained for various experimental parameters and the effect of work rate and temperature on fretting wear is evaluated.

scholarly journals 602 Development of Seismic Isolation Technology for Nuclear Power Plant (Part 2) : Component Test of Main Steam Crossover Piping

2013 ◽  
Vol 2013.21 (0) ◽  
pp. 37-38
Author(s):  
kotoyo Mizuno ◽  
Masakazu Jinbo ◽  
Hiromu Okamoto ◽  
Shinji Kosugi ◽  
Shinji Matsuoka ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Seismic Isolation ◽  
Plant Part ◽  
Component Test ◽  
Main Steam
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