Parametric Investigation on the Effect Factors for Liquid Droplet Impingement Erosion

2011 ◽  
Author(s):  
Rui Li ◽  
Hisashi Ninokata ◽  
Michitsugu Mori
Keyword(s):  
Power Plants ◽  
Erosion Rate ◽  
Structural Integrity ◽  
Volume Fraction ◽  
Liquid Droplet ◽  
Droplet Velocity ◽  
Life Extension ◽  
Droplet Impingement ◽  
Pipe Surface ◽  
Bent Pipe

Liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes of unexpected troubles occasionally occurred in the inner bent pipe surface. Evaluating the LDI erosion is an important topic of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants safety. In order to investigate the effect of various parameters, such as droplet diameter, droplet velocity and injected droplet number, on the erosion rate induced by LDI, droplet impingement under different conditions are conducted numerically by a two-phase computational approach. Considering the carrier turbulence kinetic energy attenuation due to the involved droplets, numerical simulations have been performed by using two-way vapor-droplet coupled system. This computational fluid model is built up by incompressible Reynolds Averaged Navier-Stoke equations using standard k-ε model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach, a general LDI erosion prediction procedure for bent pipe geometry has been performed to supplement an available CFD code. A correlation for the erosion rate in terms of droplet velocity, diameter and volume fraction is purposed for the engineers’ maintenance reference. Based on our computational results, comparison with an available accident data was made to prove that our methodology could be an appropriate way to simulate and predict the bent pipe wall thinning phenomena.

Numerical Analysis of Influence of Roughness of Material Surface on High-Speed Liquid Droplet Impingement

2019 ◽  
Vol 141 (3) ◽  
Author(s):  
Hirotoshi Sasaki ◽  
Yuka Iga
Keyword(s):  
Numerical Analysis ◽  
High Speed ◽  
Erosion Rate ◽  
Liquid Droplet ◽  
Material Surface ◽  
Liquid Droplets ◽  
Fluid Machinery ◽  
Droplet Impingement ◽  
Surface Liquid ◽  
Droplet Impingement Erosion

This study explains why the deep erosion pits are formed in liquid droplet impingement erosion even though the droplets uniformly impinge on the entire material surface. Liquid droplet impingement erosion occurs in fluid machinery on which droplets impinge at high speed. In the process of erosion, the material surface becomes completely roughened by erosion pits. In addition, most material surface is not completely smooth and has some degree of initial roughness from manufacturing and processing and so on. In this study, to consider the influence of the roughness on the material surface under droplet impingement, a numerical analysis of droplets impinging on the material surface with a single wedge and a single bump was conducted with changing offsets between the droplet impingement centers and the roughness centers on each a wedge bottom and a bump top. As results, two mechanisms are predicted from the present numerical results: the erosion rate accelerates and transitions from the incubation stage to the acceleration stage once roughness occurs on the material surface; the other is that deep erosion pits are formed even in the case of liquid droplets impinging uniformly on the entire material surface.

scholarly journals Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

2016 ◽  
Vol 2016 ◽  
pp. 1-10 ◽  
Author(s):  
Junxiao Zheng ◽  
Bin Zhang ◽  
Shengchun Shi ◽  
Yixue Chen
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Power Plants ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Mesh Size ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Discrete Ordinates ◽  
Reactor Pressure

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV orE>0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

Life Extension of Nuclear Power Plants: World Situation and the USA Case

2010 ◽  
pp. 50-56 ◽  
Author(s):  
Pablo T. León ◽  
Loreto Cuesta ◽  
Eduardo Serra ◽  
Luis Yagüe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Life Extension ◽  
The Usa

Comparison of the Limit Loads Between Defected Pressure Pipes With an Internal Pit and an External Pit at High Temperature

2013 ◽  
Author(s):  
Changyu Zhou ◽  
Bo Wang ◽  
Zhigang Sun ◽  
Jilin Xue ◽  
Xiaohua He
Keyword(s):  
High Temperature ◽  
Power Plants ◽  
Structural Integrity ◽  
Load Condition ◽  
Relative Length ◽  
Limit Load ◽  
Relative Depth ◽  
Corrosive Media ◽  
Pressure Pipe ◽  
Pressure Pipes

High temperature pressure pipes are widely used in power stations, nuclear power plants, and petroleum refinery, which always bear combined effects of high temperature, high pressure, and corrosive media, so the local pits are the most common volume defects in pressure pipe. Due to various reasons, the defects usually appear on the internal or external wall of pipe. In this paper, the dimensions of a defect were characterized as three dimensionless factors: relative depth, relative gradient and relative length. The main objects of study were the pipe with an internal pit and pipe with an external pit. Orthogonal array testing of three factors at four different levels was applied to analyze the sequence of the influence of three parameters. In present study, when the maximum principal strain nearby the location of the defects reaches 2%, the corresponding load is defined as the limit load, which is classified as two kinds of load type: limit pressure and limit bending moment. According to this strain criterion and isochronous stress strain data of P91 steel, the limit load of high temperature pipe with a local pit was determined by using ABAQUS. And in the same load condition of the pipe with the same dimensionless factors, the limit load of the internal defected pipe was compared with that of the external defected pipe. The results of this study can provide a reference for safety assessment and structural integrity analysis of high temperature creep pressure pipe with pit defects.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

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