Structural evaluation of a piping system subjected to thermal stratification

Piping systems in nuclear power plants are often designed for pressure, mechanical loads originating from deadweight and seismic events and operating thermal transients using the equations in the ASME Boiler and Pressure Vessel Code, Section III. In the last few decades a number of failures in piping have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. A simplified method has been developed in this work to estimate the stresses caused by the circumferential temperature distribution from thermal stratification. It has been proposed that the equation for the peak stress in the ASME Section III piping code include an additional term for thermal stratification.     

Numerical analysis of thermal striping induced high cycle thermal fatigue in a mixing tee

Thermal striping, characterized by turbulent mixing of two flow streams of different temperatures that result in temperature fluctuations of coolant near the pipe wall, is one of the main causes of thermal fatigue failure. Coolant temperature oscillations due to thermal striping are on the order of several Hz. Thermal striping high-cycle thermal fatigue that occurs at tee junctions is one of the topics that should be addressed for the life management of light water reactor (LWR) piping systems. This study focuses on numerical analyses of the temperature fluctuations and structural response of coolant piping at a mixing tee. The coolant temperature fluctuations are obtained from Large Eddy Simulations that are validated by experimental data. For the thermal stress fatigue analysis, a model is developed to identify the relative importance of various parameters affecting fatigue-cracking failure. This study shows that the temperature difference between the hot and cold fluids of a tee junction and the enhanced heat transfer coefficient due to turbulent mixing are the dominant factors of thermal fatigue failure of a tee junction.     

Experimental investigation on heat transfer characteristics in a safety injection nozzle component in PWR

Thermal fatigue is a potentially significant degradation mechanism in Nuclear Power Plants (NPP). For the fatigue analysis, the thermal load information about components must be determined firstly. In this paper, an experimental study was carried out to obtain local fluid temperatures and local heat transfer coefficients for the safety injection nozzle component in reactor coolant system (RCS). In this mixing tee component a hot jet issues into a cold cross-flow stream from an oblique pipe and the turbulent mixing of two fluids induces local cycling stresses on the adjacent piping wall. Experiments were performed using a special-made heat fluxmeter, which can measure the mixed fluid temperature close to the wall and the heat transfer coefficient between the fluid and the wall. Plexiglass and metallic 1/9-scale mockups were manufactured for flow visualization and heat transfer tests, respectively. All tests were conducted at range of 0–40 for the jet-to-cross-flow velocity ratio. The flow visualization test has obtained general pattern of the flow and identified sensitive zones in the component where the jet and cross-flow interact intensively to cause thermal fatigue more possibly. In the heat transfer test, heat fluxmeters were positioned in the wall at these sensitive zones. The measurement results of temperatures and heat transfer coefficients have been discussed in detail in the paper. These experimental results allow us improving the state of knowledge of the thermal load to be used in the industrial mixing tees in operating for long lifetime assessment and for the design in the basic Nuclear Power Plants.     

A three-dimensional method for integrated transient analysis of reactor-piping systems

A three-dimensional method for integrated hydrodynamic, structural, and thermal analyses of reactor-piping systems is presented. The hydrodynamics are analyzed in a reference frame fixed to the piping and are treated with a two-dimensional Eulerian finite-difference technique. The structural responses are calculated with a three-dimensional co-rotational finite-element methodology. Interaction between fluid and structure is accounted for by iteratively enforcing the interface boundary conditions.A thermal transient capability has been developed. A system energy equation is used to compute the coolant temperatures due to convection. A radial heat-conduction equation is employed to establish the temperature profile throughout the pipe-wall thickness. The constitutive equation used for the thermal-mechanical stress calculation is suited for a large number of materials under various loading conditions, such as those having thermal, plastic, and viscous effects. The flow surface, which defined the purely elastic regime, can be arbitrarily small; an associated flow rule is utilized for regimes of material plasticity.Three sample problems are presented to illustrate this method. The first one calculates the piping response under the seismic excitation. The second one validates the heat-conduction model. The third problem deals with a coupled hydrodynamic-structural-thermal analysis of a piping system. Results are discussed in detail.     

Xenon-ion-beam modifications of a-SiO2 and Ni/a-SiO2 layers

Modifications of a-SiO₂ films and Ni/a-SiO₂ bilayers by irradiations with 90–350 keV Xe ions have been investigated. The effects of subsequent thermal annealings in vacuum at 298–1173 K have also been studied. The analyses were performed by means of Rutherford Backscattering Spectrometry and surface profilometry. We here report on the results of ion-beam induced surface roughening and sputtering and of the noble-gas collection curves. As to the athermal ion-beam mixing at the Ni/a-SiO₂ interface, a low mixing rate in agreement with the ballistic model was observed. Only little Xe precipitation, which would indicate the presence of end-of-range spikes, occurs at the interface. Most effects were found to strongly depend on the implanted ion fluence.     

Comparison of thermo-hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17 × 17 nuclear rod bundle

Spacer grids in the nuclear fuel rod assembly maintain a constant distance between rods, secure flow passage and prevent the damage of the rod bundle from flow-induced vibration. The mixing vanes attached to the spacer grids generate vortex flows in the subchannels and enhance the heat transfer performance of the rod bundle. Various types of mixing vanes have been developed to produce cross flows between subchannels as well as vortex flows in the subchannels.The shapes of the mixing vane have been improved to generate larger turbulence and cross flow mixing. In the present study, two types of large scale vortex flow (LSVF) mixing vanes and two types of small scale vortex flow (SSVF) mixing vanes are examined. SSVF-single is conventional split type and SSVF-couple is split type with different arraying method. LSVF mixing vane has different geometry and arraying method to make large scale vortex. 17 × 17 rod bundle with eight spans of mixing vanes is simulated using the IBM 690 supercomputer. The FLUENT code and IBM supercomputer is employed to calculate the flow field and heat transfer in the subchannels.Turbulence intensities, maximum surface temperatures of the rod bundle, heat transfer coefficients and pressure drops of the four kinds of mixing vanes are compared. LSVF mixing vanes produced higher turbulence intensity and heat transfer coefficient than SSVF mixing vanes. Consequently, LSVF mixing vane increases the thermal efficiency and safety of the rod bundle.     

The turbulent mixing rate and the fluctuations of static pressure difference between adjacent subchannels in a two-phase subchannel flow

Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. In this study, flow visualization was made to investigate the mechanism of the turbulent mixing between subchannels in a two-phase flow under hydrodynamic equilibrium conditions. The test channel was a vertical multiple channel consisting of two identical rectangular subchannels, and the working fluids were air and water. It was observed in slug-churn flows that a large scale inter-subchannel liquid flow occurs in front of the nose of a large gas bubble and behind the tail when the bubble axially passes through the subchannel, and thus a high turbulent mixing rate of the liquid phase results. In order to know driving force of such a large scale inter-subchannel flow, measurement of instantaneous static pressure difference between the subchannels was also made. The result showed that there is a close relationship between the liquid phase turbulent mixing rate and the magnitude of the pressure difference fluctuations.     

Cyclic crack growth evaluation of a 20MnMoNi55 piping steel in high-oxygen reactor water

Samples of a low alloy steel piping material taken from the full scale corrosion fatigue test loop of the Heissdampfreaktor (HDR) plant have been tested at 240°C in high oxygen reactor water. The small-scale specimens (CT25) were exposed to a similar loading spectrum to that which has been used in the full-scale corrosion fatigue tests at the HDR-plant. During the autoclave tests cyclic crack growth rates were determined. Fracture surface investigations were performed not only for the laboratory test specimens but also for the fracture surface of a sample taken from the HDR test loop piping after the full scale test. In this paper the autoclave testing results and fracture surface observations are presented and compared to those obtained in the HDR piping tests.     

Characteristics of surface oscillation on high speed liquid Li jet

In the International Fusion Materials Irradiation Facility (IFMIF), high speed liquid lithium (Li) wall jet will be used as target irradiated by two deuteron beams of 125 mA at 40 MeV. To obtain knowledge of Li flow behavior, we have been studying on the surface wave characteristics experimentally using the liquid metal Li circulation loop at Osaka University. In this present study, the characteristic of surface oscillation on high speed liquid Li jet were examined. The free surface oscillation of Li flow was measured by an electro-contact probe apparatus, which detects electric contacts between a probe tip and Li surface. It was installed at 175 mm and 15 mm downstream from the nozzle exit to see influence of the initial growth of surface waves. The wave height of free surface waves was obtained from contact signal. While at 15 mm region the flow surface is very smooth covered with small waves in amplitude, the surface waves are developed sufficiently at the 175 mm. In the case of the velocity of 15 m/s, the maximum wave height reaches 4.8 mm. Heat deposition was estimated on the target back-plate with using the present statistical wave data.     

Flow mixing inside a control-rod guide tube – Experimental tests and CFD simulations

This paper covers a combined experimental and computational effort carried out at Vattenfall Research & Development AB in order to study the thermal mixing in the annular region between a top tube and a control-rod stem. The low frequency thermal fluctuations in this region can result in problems with thermal fatigue and have caused cracks in the control-rod stems of several nuclear reactors ( [Kobayashi et al., 2009] and [0070] ).The flow in the vertical annular region formed by the top tube and the control-rod stem is characterized by the mixing of hot bypass flow with cold crud-removal flow. The crud-removal flow is flowing upwards along the control-rod stem, and the warmer bypass flow is entering through eight horizontal holes positioned in the lower part of the guide tube and four holes in the upper part of the top tube, forming jets.Two full-scale models of a control rod, including the control-rod stem and the guide tube, were constructed. The first model, designed to work at atmospheric conditions, was made of Plexiglass, in order to be able to visualize the mixing process, whereas the second one was made of steel to allow for a higher temperature difference between the two flows, and the heating of the top tube.CFD simulations of the case at atmospheric conditions were also carried out.Both the experiments and the simulations showed that the mixing region between the cold crud-removal flow and the warm bypass flow is dominated by large flow structures coming from above. The process is characterized by low frequency, high amplitude temperature fluctuations. The process is basically hydrodynamic, caused by the downward transport of flow structures originated at the upper bypass inlets. The damping thermal effects through buoyancy is of secondary importance, as also the scaling analysis shows, however a slight damping of the temperature fluctuations can be seen due to natural convection due to a pre-heating of the cold crud-removal flow. The comparison between numerical and experimental results shows a rather good agreement, indicating that experiments with plant conditions are not necessary since, through the existing scaling laws and CFD calculations, the obtained results may be extrapolated to plant conditions.The problem of conjugate heat transfer has not yet been addressed experimentally since complex and difficult measurements of the heat transfer have to be carried out. These types of measurements, along with corresponding CFD calculations, constitute one of the main challenges to be dealt with in ongoing work.     

Experimental and analytical study on aerosol behavior in WIND project

The tests on fission product (FP) behavior in piping under severe accidents are being conducted in the wide range piping integrity demonstration (WIND) project at JAERI to investigate the piping integrity which may be threatened by decay heat from deposited FPs. In order to obtain the background information for future WIND experiment and to confirm analytical capabilities of the FP aerosol analysis codes, ART and VICTORIA, the FP behavior in safety relief valve (SRV) line of BWR during TQUX sequence was analyzed. The analyses showed that the mechanisms that control the FP deposition and transport agreed well between the two codes. However, the differences in models such as diffusiophoresis or turbulence, the treatment of chemical forms and aerosol mass distribution could affect the deposition in piping and, consequently, on the source terms. The WIND experimental analyses were also conducted with a three-dimensional fluiddynamic WINDFLOW, ART and an interface module to appropriately couple the fluiddynamics and FP behavior analyses. The analyses showed that the major deposition mechanism for cesium iodide (CsI) is thermophoresis which depends on the thermal gradient in gas. Accordingly, the coupling analyses were found to be essential to accurately predict the CsI deposition in piping, to which little attention has been paid in the previous studies.     

Large-Eddy Simulation study of turbulent mixing in a T-junction

A potential cause of thermal fatigue failures in energy cooling systems is identified with cyclic stresses imposed on a piping system. These are generated due to temperature changes in regions where cold and hot flows are intensively mixed together. A typical situation for such mixing appears in turbulent flow through a T-junction, which is investigated here using Large-Eddy Simulations (LES). In general, LES is well capable in capturing the mixing phenomena and accompanied turbulent flow fluctuations in a T-junction. An assessment of the accuracy of LES predictions is made for the applied Vreman subgrid-scale model through a direct comparison with the available experimental results. In particular, an estimation of the minimal mesh-resolution requirements for LES is examined on the basis of the complementary RANS simulations. This estimation is based on the characteristics turbulent scales (e.g., Taylor micro-scale) that can be computed from LES or RANS simulations.     

From discovery to recognition of periodic large scale vortices in rod bundles as source of natural mixing between subchannels—A review

The mixing of cooling fluid in rod bundles from one subchannel to another through the gaps between the rods reduces the temperature differences in the coolant as well as along the perimeter of the rods. The phenomenon of natural mixing was first intensively investigated in the 1960s and remains a topic of research up to the present time. The paper describes the main stations on the way to understand the nature of the flow in rod bundles and generally in compound channels with the focus on work performed at Research Center Karlsruhe (FZK).¹Earlier, it was noticed that the mixing rates where higher than could be accounted for by turbulent diffusion alone. For more than 20 years attempts were made to prove experimentally and by code application that secondary flows could account for the measured mixing rates, although the measured secondary flow velocities were much too low. Measurements of the turbulence structure by hot wire anemometry confirmed the existence of cyclic flow pulsations, which had been postulated earlier on the basis of thermocouple measurements. More sophisticated hot wire measurements revealed the nature of these pulsations as periodic, coupled to gap width and Reynolds number. Finally, the extension of the investigation to other compound channel types and flow visualization revealed the true nature of the mixing process as a vortex train moving along the gap between rods or in the narrow part of a compound channel. These findings have been confirmed by LES calculations. Based on these results CFD codes with improved turbulence models calculated successfully the flow in rod bundles including the macroscopic oscillations.     

Experimental investigation on similarity between velocity and density profiles in density-stratified countercurrent flow in reactor horizontal leg

A density-stratified countercurrent flow was investigated to obtain data necessary to develop a physical model on a thermally stratified flow in a horizontal leg of a pressurized water reactor (PWR). The experiments were conducted at atmospheric pressure and temperature using fresh water and NaCl solution with a non-dimensional density ratio of up to 1.2. The emphasis was placed on measurements of velocity and concentration profiles near the interface between the two fluid layers. Measured mean velocity and concentration profiles were fitted consistently using the Monin–Obukhov similarity theory, which are well-known outcomes for stratified turbulent shear flow. The interfacial friction and entrainment coefficients obtained from the fitted profiles agreed well with existing results in literature, confirming the applicability of the Monin–Obukhov theory. Furthermore, a new empirical correlation was proposed for the prediction of a mixing layer thickness.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

Fluid mixing and flow distribution in the reactor circuit, measurement data base

Experimental investigations and computational fluid dynamics (CFD) calculations on coolant mixing in pressurised water reactors (PWR) have been performed within the EC project FLOMIX-R. The project aims at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. Measurement data from a set of mixing experiments have been gained by using advanced measurement techniques with enhanced resolution in time and space. Slug mixing tests simulating the start-up of the first main circulation pump are performed with two 1:5 scaled facilities: the Rossendorf Coolant Mixing model ROCOM and the Vattenfall test facility. Additional data on slug mixing in a VVER-1000 type reactor have been gained at a 1:5 scaled metal mock-up at EDO Gidropress. Experimental results on buoyancy driven mixing of fluids with density differences have been obtained at ROCOM and the Fortum PTS test facility.Concerning mixing phenomena of interest for operational issues and thermal fatigue, flow distribution data available from commissioning tests at PWRs and VVER are used together with the data from the ROCOM facility as a basis for the flow distribution studies.In the paper, the experiments performed are described, results of the mixing experiments are shown and discussed. Efforts on computational fluid dynamics codes validation on selected mixing tests applying Best Practice Guidelines in code validation will be reported about in a separate paper.     

Development of Ultrasonic Testing Equipment Incorporating Electromagnetic Acoustic Transducer

An ultrasonic testing equipment for use in in-service inspection of nuclear power plant piping has been developed, which comprises an angle-beam electromagnetic acoustic transducer mounted on a vehicle for scanning the piping surface to be inspected. The transducer functions without direct contact with the piping surface through couplant, and the vehicle does not require a guide track installed on the piping surface, being equipped with magnetic wheels that adhere to the piping material, permitting it to travel along the circumferential weld joint of a carbon steel pipe. The equipment thus dispenses with the laborious manual work involved in preparing the piping for inspection, such as removal of protective coating, surface polishing and installation of guide track and thereby considerably reduces the duration of inspection. The functioning principle and structural features of the transducer and vehicle are described, together with the results of trial operation of a prototype unit, which proved a 1 mm deep notch cut on a test piece of 25 mm thick carbon steel plate to be locatable with an accuracy of ±2 mm.     

Development of flow-induced vibration evaluation methodology for large-diameter piping with elbow in Japan sodium-cooled fast reactor

This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in Japan sodium-cooled fast reactor, with particularly emphasis on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. The approach to the methodology development was defined: experiment-based methodology and simulation-based one as well as extrapolation logic to the reactor condition based on no dependency on Reynolds number in the high Reynolds number range from the experimental results. Experimental efforts have been made using 1/3-scale single-elbow test sections for the hot-leg piping as the main activity. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced pressure fluctuations onto the pipe though a slight deformation of flow separation was observed. Numerical results under different Reynolds number conditions appear in this paper using the unsteady Reynolds Averaged Navier Stokes equation approach, indicating its applicability to the hot-leg piping experiments.     

探针组分对紫外光固化材料表面性能的影响

光固化快速成型技术（Stereolithography，SL）采用材料逐层固化累加法来形成三维实体模型，因此固化材料的表面性质对层与层之间的结合力有着非常大的影响。为了研究环氧丙烯酸酯类紫外光固化材料所形成表面的性质及影响因素，通过使用探针组分，对固化涂层进行接触角的测定和表面能的计算，并用衰减全反射-傅立叶变换红外光谱（Attenuated total reflectance-Fourier transforminfrared，ATR-FTIR）对固化膜的表面组成进行了表征。结果表明：固化过程中组分发生了迁移，其含量由表及里存在分布不均现象，最终对固化层的表面能产生了影响，且质量分数为0．01％的探针组分含量就能使表面能的变化显得相当敏感（Deuchem 467使涂层表面能由51．32mJ m^-2骤降至35．05mJm^2）.不同结构的探针组分迁移程度有差异，使涂层表面能发生了不同的改变。同时，组分的迁移程度也与涂层的两种接触介质（固化气氛与基材）的性质有关，引起涂层气氛面和基材面表面能的较大差异。     

The steamhammer problem: dynamic shock loading of critical reactor piping systems

In the course of both pre-operational testing and power operation of commercial nuclear power plants, relatively large amplitude transient vibrations of steam piping systems have been experienced with damage to the piping supports in at least one recent case. These transient vibrations result from ‘steamhammer’ or dynamic shock loading induced by pressure and momentum transient conditions generated in the piping by sudden changes to the flow conditions, such as are produced by sudden valve opening or closure. In particular, vibrations have been experienced in by-pass and discharge lines as a result of relief valve discharge, and in main steam lines as a result of sudden main stop valve closure. Piping in both BWR and PWR reactor systems has been found to be susceptible to these conditions.This paper is concerned with the evaluation of the pressure and momentum transients resulting from sudden valve operation, and the determination of the dynamic response of the piping to the induced transient loading. The characteristics of the transient conditions existing immediately following both sudden valve opening and closure as encountered in BWR and PWR plants are discussed. The procedures used to calculate the transient time history functions are outlined. The derivation of the loading induced in the piping by the pressure and momentum transients is discussed and the determination of the dynamic response of the piping is presented. The procedures described in the paper are illustrated by actual examples from BWR and PWR plants, and the significance of steamhammer effects relative to other loading conditions is discussed.