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This paper describes the experimental study on the liquid droplet impingement erosion of metal materials to understand the influence of material hardness on the erosion rate. The experiment is carried out using a water spray jet apparatus with a condition of relatively thin liquid film thickness. The metal materials tested are pure aluminum, aluminum alloy, brass, mild steel, carbon steel and stainless steel. The liquid droplets considered are 30 ± 5 μm in volume average diameter of water, which is the same order of droplet diameter in the actual pipeline in nuclear/fossil power plants. In order to understand the influence of material hardness on the liquid droplet impingement erosion, the scanning electron microscope (SEM) observation on the eroded surface and the measurement of erosion rate are carried out in the terminal stage of erosion. The experimental results indicate that the erosion rates are expressed by the droplet velocity, volume flux, Vickers hardness and the liquid film thickness, which are fundamentals of the liquid droplet impingement erosion. The empirical formula shows that the power index for droplet velocity dependency is found to be 7 with a scattering from 5 to 9 depending on the materials, while the power index for Vickers hardness dependency is found as −4.5.

Flow-induced acoustic resonances in piping with closed side branches or T-junctions are one of the causes of severe structural vibrations, which sometimes cause fatigue damage to piping and components in a power plant and many engineering applications. In this paper, on the basis of the results of steam flow experiments and calculations, the effects of the liquid phase on the flow-induced acoustic resonance at closed side branches in the steam flow piping of BWRs are described, and some suggestions for the steam piping design of BWRs are also given. The liquid phase in a steam flow forms droplets or liquid film, which may affect the amplitude, frequency and critical Strouhal number of the resonance. From the results of wet steam experiments and CFD calculations, we have found that in some cases the wetness of the steam flow may decrease the resonant amplitude and change the frequency owing to the interaction of the vortex generation or damping by the existence of the liquid film and droplets. Therefore, for the wet steam piping design of BWR, some suggestions for taking these effects into consideration, under actual BWR steam conditions are described.

A steam control valve is used to control the flow from the steam generator to the steam turbine in thermal and nuclear power plants. During startup and shutdown of the plant, the steam control valve is operated under a partial flow conditions. In such conditions, the valve opening is small and the pressure deference across the valve is large. As a result, the flow downstream of the valve is composed of separated unsteady transonic jets. Such flow patterns often cause undesirable large unsteady fluid force on the valve head and downstream pipe system. In the present study, various flow patterns are investigated in order to understand the characteristics of the unsteady flow around the valve. Experiments are carried out with simplified two-dimensional valve models. Two-dimensional unsteady flow simulations are conducted in order to understand the experimental results in detail. Scale effects on the flow characteristics are also examined. Results show three types of oscillating flow pattern and three types of static flow patterns.

A steam control valve is used to control the steam flow from a steam generator to a turbine in power plants. The flow field in the steam control valve is usually complicated and transonic. A flow-induced vibration can occur in the valve and may cause unexpected shutdown of the plant. An experimental study is conducted to clarify the characteristics of the flow in the steam control valve. In the present study, scale effects on the condition of occurrence of various flow patterns are investigated. Experimental results show various flow patterns including new patterns which have not been reported. Additionally, even under the same operation condition, different flow patterns are observed under a certain range of condition. This effect is found to be comparable with the scale effect.

In nuclear power plants, fluid structure interactions (FSI) occurring in component systems can cause excessive forces or stresses to the structures resulting in mechanical damages that may eventually threaten the structural integrity. FSI in the guidelines includes flow-induced vibration, water hammer, and pipewhip. It can also include movement, deformation, or fracture of equipments by tsunami etc. They can be issues of design and maintenance. Authors cannot find complete guidelines to correspond to the FSI phenomena which can be important in the design and maintenance of nuclear power plants. Based on the background, International Atomic Energy Agency (IAEA) has drafted guidelines on FSI. This paper summarizes general description of FSI as well as design and maintenance against FSI.Copyright © 2014 by ASME

Structural strengths of the piping and components in NPPs have been designed with seismic margin. They are classified seismically S, B and C class in terms of the influence rate to nuclear safety. For the highest seismic class (Class S) equipment, it is clarified that they have enough seismic margins against design seismic conditions by shaking table tests or numerical simulations. However, for the lower seismic class (Class B and C) equipment, their seismic margins have not been clarified quantitatively. In this paper, in order to evaluate seismic robustness of the lower seismic class equipment with no clarification of seismic margin, seismic influences of the lower seismic class equipment in NPPs damaged by actual large earthquakes have been surveyed and sorted as a database, and the integrity of the lower class equipment have been discussed. Seismic effects on 24 plants damaged by the recent large 6-earthquakes are surveyed, sorted as a database, and investigated. As a result, a total of 29 cases of function deterioration or loss were observed. Considering the total number of components and piping, the frequency of those cases in class B and C components and piping was low. And also, as it is found there are a few cases of degradation or loss of function in the equipment installed on the bedrock or in the buildings.

Thermal-hydraulics governs mechanical load and environmental condition of structural materials, thus a quantitative understanding of thermal-hydraulic behavior is of real importance to a correct estimate of many kinds of plant aging processes. The current status of thermal-hydraulics for prediction of plant aging is summarized and subjects for future research are discussed.

Pipe wall thinning is a one of the major degradation mechanisms in aged nuclear power plants (NPPs). In Japan, the seismic safety of wall thinning piping system during earthquake must be evaluated in aged NPPs. Seismic safety of piping systems with wall thinning had been investigated by other researchers using shaking table tests of reduced scale and numerical analyses. However, there exist the limitations such as the scale effect of pipe model for shaking table tests and the limit of the evaluation for numerical analysis concerning the criteria of pipe integrity. By the way, elbow can be one of the most important elements to evaluate the seismic safety of piping system. So, in order to evaluate seismic safety of piping systems with wall thinning elbow, hybrid tests have been conducted, in which the seismic response of the whole piping system is treated as a numerical model, and the real piping is used only for the element on which the transformation and damage locally concentrated. The through-wall crack only occurred in the case of a uniform thinning model although cracks didn’t penetrate in the non thinning model and the local thinning model. In the experimental condition, the failure mode of wall thinning elbow under seismic loadings had been low cycle fatigue, and effectiveness of this evaluation method has been demonstrated.

Flow accelerated corrosion (FAC) is one of the pipe wall thinning phenomena. There are a few prediction equations for FAC under the two-phase flow condition, but their accuracy is uncertain. In this study, these equations are summarized. A new prediction equation is proposed, in which the liquid film velocity near the wall is taken into account. Also, FAC experiments with water-steam two-phase flow are performed. The thinning rate is predicted by calculation and compared with the experimental value.Copyright © 2013 by ASME

Flow accelerated corrosion (FAC) and liquid droplet impingement erosion (LDI) are the main pipe wall thinning phenomena in piping system of power plants in Japan. Authors have promoted the development of prediction method to evaluate local thinning trend by FAC/LDI. To apply the method to pipe wall thinning management in power plants, it is required to be transformed into practical tools for easy usage. In Japan, discussion is being made to considerate the introduction of prediction tools into wall thinning management based on wall thickness measurement at present. Authors have simplified their FAC/LDI models to predict wall thinning trend one-dimensionally along piping layout, and applied to actual thinning data of power plants. With PWR’s FAC data and BWR’s LDI data, maximum thinning rate for each pipe elements were roughly predictable with considerable accuracy. Especially for high thinning rate data, which is important in plant management, the model was able to evaluate within the factor of 2. By installing this model, prediction software “FALSET” was developed, equipped with practical functions for the management. With the further verification and improvement of each function, there are prospects for this software to be utilized as a management tool in power plants.

Flow accelerated corrosion (FAC) of the carbon steel is one of the most important subjects in the coolant systems of the power plants. FAC is influenced by the composition of the material, the flow condition, temperature, and the water chemistry conditions. It is considered that the solubility of iron (Fe) is the most important factor in the water chemistry parameters affecting FAC. In the present study, the effects of temperature and pH on the Fe solubility were evaluated in consideration of the hydrolysis reactions of the ferrous iron, the dissolution equilibria of Fe3O4, FeO, and Fe(OH)2, and the charge balance. The correlation between the Fe solubility and the FAC behavior was discussed. It has been suggested that the product of the Fe solubility equilibrated with Fe3O4 and the mass transfer coefficient can explain the temperature and pH dependence of FAC. These results indicate the presence of the magnetite on the surface of the carbon steel. Diffusion of the Fe from the saturated layer to the bulk solution determines the FAC rate from the water chemistry aspect.

For seismic safety evaluation of piping system with local thinning surface by liquid droplet impingement erosion (LDI), hybrid seismic tests were conducted to the piping with a locally-thinned elbow. In this paper, a method for predicting the thinning shape by LDI on the elbow is developed. To determine the thinning shape by LDI, droplet behavior at the elbow is calculated for various flow conditions and geometries. With the calculation of the collision point and velocity for each droplet, collision frequency and average collision velocity on the elbow are estimated. Then, the thinning shape on the elbow is determined with the relationship between the flow conditions and thinning rate. Finally, the evaluated thinning shape is compared with an actual LDI case for the validation of the method.Copyright © 2011 by ASME

It is well known that the wetness of steam flow sometimes causes measurement errors of the steam flow meter. However, it is difficult to clarify a particular error quantitatively in actual plants and factories, and thus far, there has been no established method for estimating the error caused by the wetness of steam flow. Therefore, wet steam flow rate measurement experiments were conducted to clarify the measurement error caused by the wetness of steam flow in a plant and a factory. In this study, as the first step, the orifice flow meter was applied because it is the main flow meter in actual plants. Experiments were conducted with the steam flow apparatus by changing the flow rate, pressure and wetness. As a result, the correlation between the measurement error and the flow condition was clarified. Moreover, for the correction of the error, a new correction method was applied and was confirmed to be better than existing methods now being used.Copyright © 2011 by ASME

Newly developed wet steam flowrate calibration facility is introduced. It has a closed loop in which boilers generate a steam flow up to 800 kg/h. Steam flow of known wetness up to 12% is generated by cooling down a dry steam flow by a heat exchanger. The wetness is calculated from the enthalpy the heat exchanger draws from the dry steam flow. Analysis of the facility performance, calibration results of an orifice flowmeter calibration, and uncertainty analysis are described.Copyright © 2011 by ASME

The power uprating of the nuclear power plant is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. We have started to develop the evaluation methods of pressure fluctuations at the dryer in BWR-5, which are the main type of BWR in Japan. In this study, to investigate the fluctuating pressure at rated power condition, we conducted steam flow simulations in actual main steam piping geometry and conditions. As a result, small amplitude resonance was observed in rated power condition.Copyright © 2010 by ASME

Flow Accelerated Corrosion (FAC) is well known as a complex phenomana of hyraulics and electro-chemicals. Among the two, this study focused on the hydraulic factors affecting FAC. FAC experiments with small rectangular flow duct were conducted in PWR condensate condition. Flow field for the experiment was calculated with numerical simulation using LES (Large Eddy Simulation) turbulence model. From both experimental and numerical results, new model of mass transfer coefficient, as the essential parameters of hydraulics, was proposed considering local turbulent velocity, so as to evaluate the effect of eccentric flow on FAC. To verify the applicability of the model, FAC plant data of actual PWR (Pressurized Water Reactor) condensate line (146 degC) and BWR (Boiling Water Reactor) condensate line (35 degC) were referred. Mass transfer coefficients for each pipe lines were calculated from flow numerical analysis. The new proposed model showed good correlation with the data of FAC thinnng rate, and its applicability was confirmed. In addtion, comparing the two plant cases, electro-chemical effect could be estimated as a similar level, which suggests the possibility of low-temperature FAC in BWR condensate lines.Copyright © 2009 by ASME

Liquid droplet impediment erosion (LDI) is a pipe wall thinning phenomena that a droplet accelerated by steam flow attacks pipe surface. It is difficult to evaluate LDI behavior in the system of NPP. Therefore, in current pipe wall thinning management, the pipe element is replaced conservatively when it is recognized to become thinner by pipe wall thinning measurement result. If LDI behavior at each point of the system can be estimated, it is possible to measure the pipe wall thickness and replace the pipe element at an appropriate time, in accordance with reasonable pipe wall measurement schedule. CRIEPI has constructed the evaluation system for LDI by using flow dynamics. This paper is concerned with the following items: • the applicability of the LDI evaluation system by a comparative verification between pipe wall thinning rate calculated by LDI system and that by actual measurement result; • establishment of remaining life assessment method of pipe in LDI environment by using LDI model.Copyright © 2009 by ASME

In some cases, a steam control valve in a power plant causes a large vibration of the piping system under partial valve opening. For rationalization of maintenance and management of a plant, it is favorable to optimize the valve geometry to prevent such vibration. However, it is difficult to understand the flow characteristics in detail only from experiments because the flow around a valve has a complex 3D structure and becomes supersonic (M>1). Therefore, it is useful to combine experiments and CFD (Computational Fluid Dynamics) for the clarification of the cause of vibration and optimization of valve geometry. In previous researches involving experiment and CFD calculation using “MATIS” code, we found that an asymmetric flow attached to the valve body (named “valve-attached flow”) occurs and pressure increases where the valve-attached flow collides with the flow from the opposite side under the middle opening condition. This high-pressure region rotates circumferentially (named “rotating pressure fluctuation”) and causes cyclic side load on the valve body. However, because we assumed the valve support is rigid, we cannot clarify the interaction between the rotating pressure fluctuation and the valve vibration when the valve stiffness is small. Thus, in this paper, we conducted flow-induced vibration experiments on a valve with a very weak support and investigated the characteristics of the vibration mode under the middle-opening condition. As a result, under the specific lift condition of the region where rotating pressure fluctuation occurs, lock-in phenomena between the rotating pressure fluctuation and the valve vibration occur and large-amplitude vibration can be seen.Copyright © 2005 by ASME

A steam control valve causes vibrations of piping when the opening is in the middle condition. For the rationalization of maintenance and management in a plant, valves should be improved, but it is difficult to understand the flow characteristics in detail experimentally because the flow around the valve has a complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of vibration and to develop improvements both experimentally and through CFD (Computational Fluid Dynamics) calculation. First, small-scale air experiments are performed. In this experiment, a spike-type pressure fluctuation that rotates circumferentially is observed in a middle valve opening. This fluctuation sometimes changes its rotation direction suddenly, or sometimes stays at the same position. Because of this randomness, FFT analysis cannot reproduce this peak frequency. However, another peak caused by the 1st-mode resonance in the pipe-diameter direction can be seen. After some experiments, a high-precision 3D CFD code for compressible flow, “MATIS”, was developed. This code can calculate an unsteady 3D compressible flow steadily and accurately, with TVD scheme, LU-SGS implicit scheme and inner iterative calculation. This code also includes LES turbulence model for compressible fluid in order to reproduce turbulence flow accurately. To validate this code, first, some benchmark tests such as karman vortex calculation and the calculation of a detached shock position on a sphere are performed. After benchmark tests, CFD calculation of the valve is performed under several valve opening conditions. The mesh number is about 600 thousand and y+ is about 3. CFD results agree well with those of the experiment qualitatively and quantitatively. Therefore, the validation of MATIS is confirmed. Detailed flow characteristics around the valve in a middle valve opening are investigated. Under this condition, the jet after the throat travels the valve (named “attached flow”) and strikes the jet coming from the opposite side. This phenomenon generates a high-pressure region and a pair of vortices; this region rotates circumferentially and causes cyclic side load on the valve body. CFD calculation clarifies that a circumferentially propagating spike-type pressure fluctuation is caused by attached flow along the valve, and this is though to be the cause of vibration.