Abstract The prediction of gas and liquid leak rate through packed stuffing boxes subjected to gas flow is a subject of very few studies in the literature. For better prediction of leakage, the change of porosity with length due to the non-uniform axial stress must be accounted for. There are few theoretical models on the prediction of leak rates in packing rings with capillary models. However, a model that incorporates the change of the capillary area with stress gives a better prediction. In this paper, the first slip flow model is used to predict gas and liquid flow considering the straight capillaries and capillaries having an area dependent on the axial stress in the packing rings. An approach that uses an analytical-computational methodology based on the number and the size of pores obtained experimentally is adopted to predict gas and liquid leak rates in uniform and non-uniform compressed yarned packings. The Navier-Stokes equations associated with slip boundary condition at the wall are used to predict leakage. Experimental tests with helium, argon, nitrogen and air for gazes and water and kerosene for liquids will be used to validate the models. The porosity parameters characterization will be conducted experimentally with helium at a reference gas at different gland stresses and pressures.

Abstract A challenge for owners operating nuclear power plants and spent fuel canisters is to provide sound technical basis for the safety and security of long-term operation and storage, respectively. Two code cases N-729-6 and N-770-5 provide performance requirements for mitigation of Primary Water Stress Corrosion Cracking (PWSCC) in ASME Section XI and Code Case N-860 “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems” is under development in ASME Section XI. The primary degradation concern being addressed in N-860 is Chloride Induced Stress Corrosion Cracking (CISCC). ASME Section III has formed two Task Groups with the intent to reduce PWSCC and CISCC in operation by mitigating residual stresses during new construction of components. ASME Section III formed Task Group Weld Residual Stress (TG-WRS) to incorporate lessons learned into the Code from operating plants regarding their experience with stress corrosion cracking. An example of expected guidance from TG-WRS is to consider residual stresses after weld repair of a susceptible material when the repair surface is exposed to the wetted environment. To address this concern, one form of mitigation could be the use of SSI which has been used in operating plants. However, there is a gap because no acceptance or performance criteria exists for mitigation of SCC by surface stress improvement for ASME Section III components. The executive committee of ASME requested that a new Task Group named “Advanced Surface Stress Improvement Technology” (ASSIT) be formed by the Korean International Working Group (KIWG) and Materials Fabrication and Examination (MF&E) to address this gap by developing a code case for mitigation of PWSCC and CISCC for new ASME Section III components by SSI techniques. This paper will discuss the background and current approach being taken by Task Group ASSIT in developing SSI acceptance or performance criteria for mitigation of SCC. The necessary technical reports supporting this code case are also under development as part of joint projects between Doosan Heavy Industries and Construction (DOOSAN), Electric Power Research Institute (EPRI), and Sun Moon University (SMU). Interim reports summarizing the status of these joint projects will be presented as separate papers in the same session.

Abstract The Primary Water Stress Corrosion Cracking causes existing ASME BPV Code Section XI inspection and repair methodology to need enhancement. This forms the basis for Code Case N-770 and revisions through 6. The paper includes the bases for examination requirements, examination schedule, and repair criteria.

Abstract It is necessary for nuclear power plant operation and spent fuel canisters to provide a sound technical basis for the safety and security of long-term operation and storage respectively. A new code case for mitigation of Primary Water Stress Corrosion Cracking (PWSCC) and Chloride Induced Stress Corrosion Cracking (CISCC) in Section III components by using an advanced surface stress improvement technology (ASSIT) is being developed by Task Group ASSIT which is one of the task groups under the ASME (The American Society of Mechanical Engineers). The necessary technical reports supporting this code case are being developed as part of joint research projects conducted by Doosan Heavy Industries and Construction (DOOSAN), Electric Power Research Institute (EPRI) and Sun Moon University (SMU). A well-known approach to prevent PWSCC and CISCC are to be performed using materials resistant to PWSCC and CISCC. The objective is to eliminate residual tensile stresses, or to induce compressive residual stress using ASSIT methods such as laser peening, water jet/cavitation peening, ultrasonic peening and ultrasonic nanocrystal surface modification (UNSM). Performance and measurement criteria for mitigation of PWSCC by ASSIT will be established based on the magnitude of surface stress and depth of compressive residual stress, sustainability, inspectability and lack of adverse effects. Additionally, for mitigation of CISCC by ASSIT, the evaluation of chloride induced corrosion pitting, the depth and density of corrosion pits and stress corrosion crack initiation and growth under chloride salt chemistry conditions are also being examined. This paper explains the approach, and progress of testing and analysis. The results and details from testing and analysis will be presented in a future PVP paper upon completion.

Abstract Future code development to adopt a risk-informed design methodology will require improved accuracy of fatigue initiation predictions. The ability to account for through wall strain gradients in plant components, particularly over the first 3 mm of wall thickness is one area where conservatism can be reduced. This is due to extant design fatigue curves being derived from strain controlled membrane loading tests where the 25% load drop definition for end of test equates to approximately a 3 mm crack. Being able to define initiation fatigue curves for much shorter crack depths would enable fatigue crack growth methods to then predict the additional cycles, taking into consideration the strain gradient, required before the defined end of life crack size is reached. The R5 procedure provides a method to adjust an existing Stress-Life (SN) curve representing an initiation crack depth, to a smaller depth. This method was developed for materials at higher temperature and for a CO 2 cooling environment, thus its validity was uncertain for application to a Pressurised Water Reactor (PWR) plant. This paper details the development of best estimate and design basis SN curves and environmental fatigue enhancement factors (F en ) for crack initiation to a depth of 250 μm. It is concluded that the general methodology in R5 was found, through this work, to adequately describe fatigue initiation lives for stainless steel in a PWR environment when augmented with a crack size dependent F en equation and with modified coefficients determined through regression. The resulting method is similar to R5 in that an SN curve can be derived for any crack size, however the current data set only provides validation down to a crack size of 250 μm, as striations at shorter depths were not visible with existing methods.

Abstract For reactor pressure vessels (RPVs) of pressurized water reactors, temperature of the coolant water in the emergency core cooling system (ECCS) may influence the structural integrity of the RPV during pressurized thermal shock (PTS) events. By focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. The analysis results show that the failure probability of RPV decreased dramatically when the coolant temperature in accumulator as well as in the high- and low-pressure injection systems (HPI/LPI) was increased, although the increase in coolant temperature in the HPI/LPI only did not lead to a decrease in the failure probability.

Abstract Since 1982 the nuclear industry has employed weld overlay repairs to address intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWR) and primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR). The American Society of Mechanical Engineers (ASME) has created several documents to provide rules and guidelines for weld overlay repair of nuclear components that have experienced stress corrosion cracking (SCC). These documents include ASME Code Case N-504-4 and ASME Section XI, Nonmandatory Appendix Q which specifically address weld overlay repair of stainless steel components. Recently, stainless steel components that have experienced thermal fatigue cracking at the inner diameter surfaces have been repaired with weld overlays using the methodology of Case N-504-4 and Appendix Q. This repair technique is appropriate to address thermal fatigue cracking because it provides structural reinforcement to the affected location and places the inside diameter (ID) surface into compression preventing, or significantly reducing, further flaw growth. However, the rules of Case N-504-4 and Appendix Q were not specifically written to address thermal fatigue cracking as the primary cause and may not adequately address design, analysis and examination requirements when thermal fatigue is the active mechanism because it is very different in nature than SCC. For example, SCC is driven by a combination of steady state operating stresses, residual stresses from welding and fabrication processes, and temperature, whereas thermal fatigue is driven by thermal stress cycles resulting from fluid thermal cycling or stratification. The source of the thermal events that result in cracking may not be as well understood or predicable as SCC degradation. Therefore, alternative rules are needed for repair of piping damaged by thermal fatigue. This paper provides a technical basis for weld overlay repair of components that have experienced thermal fatigue cracking. It addresses design, analysis and examination requirements considering the nature of thermal fatigue in nuclear piping systems. The basis begins with the industry accepted rules of Case N-504-4 and Appendix Q and discusses the appropriate modifications needed to address thermal fatigue cracking. These modifications include removing restrictions such as the delta ferrite limit that is only applicable to address SCC, and enhancements to the examination requirements to ensure that the repaired location is adequately monitored throughout the remaining service life of the plant. The purpose of this paper is to outline the basis for a new ASME Code Case that will contain the appropriate rules for weld overlay repair of Class 1, 2 and 3 stainless steel piping degraded by thermal fatigue cracking. The new Case is currently in draft form and the requirements and specific details are still evolving. Thus, it is envisioned that this technical basis will be revised to include updates and revisions to the Case.

Abstract For several decades pressurized water reactors have experienced Primary Water Stress Corrosion Cracking (PWSCC) within Alloy 600 components and welds. The nuclear industry has developed several methods for mitigation of PWSCC to prevent costly repairs to pressurized water reactor (PWR) components including surface stress improvement by peening. Laser shock peening (LSP) is one method to effectively place the surface of a PWSCC susceptible component into compression and significantly reduce the potential for crack initiation during future operation. The Material Reliability Program (MRP) has issued MRP-335, which provides guidelines for effective mitigation of reactor vessel heads and nozzles constructed of Alloy 600 material. In addition, ASME Code Case N-729-6 provides performance requirements for peening processes applied to reactor vessel head penetrations in order to prevent degradation and take advantage of inspection relief, which will reduce operating costs for nuclear plants. LSP Technologies, Inc. (LSPT) has developed and utilized a proprietary LSP system called the Procudo ® 200 Laser Peening System. System specifications are laser energy of 10 J, pulse width of 20 ns, and repetition rate of 20 Hz. Scalable processing intensity is provided through automated focusing optics control. For the presented work, power densities of 4 to 9.5 GW/cm 2 and spot sizes of nominally 2 mm were selected. This system has been used effectively in many non-nuclear industries including aerospace, power generation, automotive, and oil and gas. The Procudo ® 200 Laser Peening System will be used to process reactor vessel heads in the United States for mitigation of PWSCC. The Procudo ® 200 Laser Peening System is a versatile and portable system that can be deployed in many variations. This paper presents test results used to evaluate the effectiveness of the Procudo ® 200 Laser Peening System on Alloy 600 material and welds. As a part of the qualification process, testing was performed to demonstrate compliance with industry requirements. The test results include surface stress measurements on laser peened Alloy 600, and Alloy 182 coupons using x-ray diffraction (XRD) and crack compliance (slitting) stress measurement techniques. The test results are compared to stress criteria developed based on the performance requirements documented in MRP-335 and Code Case N-729-6. Other test results include surface roughness measurements and percent of cold work induced by the peening process. The test results demonstrate the ability of the LSP process to induce the level and depth of compression required for mitigation of PWSCC and that the process does not result in adverse conditions within the material.

Abstract During the solid fluidization exploitation of shallow non-diagenetic NGHs (Natural Gas Hydrates) in the deep-water, hydrates together with mineral sand, natural gas, seawater and drilling fluids flow in the production pipeline. Natural gas released from hydrates during the process of solid fluidization will reform hydrates under the suitable conditions. Therefore, research on the formation and dissociation of methane hydrates in the presence of fine-grain sands is of great significance for ensuring the flow assurance of solid fluidization exploitation of shallow non-diagenetic NGHs in the deep-water field. In this paper, a high-pressure autoclave was used to carry out the experiments of hydrate formation and dissociation under different initial pressures and particle sizes of the fine-grain sand, for investigating into the hydrate induction time, formation amount, rate and dissociation affected by the presence of the fine-grain sand. Results indicated that hydrate formation kinetics in the presence of fine-grain sand was supposed to be also affected by mass/heat transfer, thermodynamics and kinetics. The fine-grain sand would be dispersed in the water phase under the effect of buoyancy, gravity and shearing force. Besides, the fine-grain sand at the gas-water interface would hinder the mass transfer of the methane gas into the water, inhibiting the nucleation of the hydrates, which was more obviously at the lower pressure. When the driving force for hydrate formation was larger, hydrate formation amount increased with the decrease of the particle size of the fine-grain sand. However, hydrate formation amount decreased with the decrease of the particle size of the fine-grain sand when the driving force for hydrate formation was lower. The average growth rate in the presence of fine-grain sand with 2.9 μm was larger than that of 9.9 μm. However, hydrates grew rapidly and subsequently tended to grow at a lower rate in the presence of fine-grain sand with 2.9 μm at 8.0 MPa initial pressure, which was assumed to be affected by the unconverted water wrapped inside the hydrate shell. The changing trends of gas emission during the dissociation process between the sand-containing system and the pure water system were nearly the same. The amount of gas emission reached a peak value within 15 minutes and then tended to stabilize. The difference in the amount of gas emission mainly depended on the formation amount before hydrate dissociation. Hydrates grew rapidly once methane hydrates nucleated in the presence of the fine-grain sand at the lower pressure, which would increase the plugging risk during the process of the solid fluidization exploitation. Further study of the fine-grain sand on flow assurance during hydrate dissociation process should be done in the future. The results of this paper provided an important theoretical basis and technical support for reducing the risk in the process of the solid fluidization exploitation of shallow non-diagenetic NGHs in the deep-water field.

Abstract The long period velocity pulse is recognized as one of the characteristics of near-fault ground motions, and hence the response of vibration modes with lower frequencies will be amplified owing to the resonant effect. In general, the sloshing frequency of storage liquid is low and the period is similar to the pulse period of near-fault ground motions. Compared to the far-field ground motions, the induced sloshing height will be amplified by the near-fault ground motions. Therefore, it is worth paying attention to the resonant effect of near-fault ground motions on the sloshing mode of storage liquid in tanks. An experiment was implemented to study the resonant response of sloshing mode. The purpose of this experiment is to estimate the slosh height and the associated total volume of water splashing out of the tank under near-fault ground motions, and also to determine the relationship between the resonant response and the input velocity pulse. This paper aims to describe the test plan in detail, and it consists of (1) design of the scaled storage tank and water depth, (2) selection and processing of the input motions including the original near-fault ground motions, extracted velocity pulse or extracted bandpass signals for resonance analysis, and also impulse motion for free vibration, (3) setup of measure instrument, and (4) the experimental procedures as well. Preliminary analysis results are compared with the code-specified values that is determined by the industrial standards and guidelines for general seismic conditions. It is noted that the proposed prediction equation can be applied to the seismic design and evaluation of spent fuel pool in nuclear power plants.

Abstract Nuclear power plant spent fuels are initially stored in the spent fuel pool. Then, the water cooled fuels are transferred in a concrete or steel cask and transported outside of the Fuel Handling Building (FHB) or the Reactor Building (RB) for long term on site storage. The spent fuel casks are typically stored on a slab-on-grade pad. The slab-on-grade pad is designed according to the U.S. Nuclear Regulatory Commission NUREG-1536 and NUREG-1567. The two Standard Review Plans provide guidance to the regulators for the review of cask storage system license application. The ISFSI pad analysis and design have to consider various loading conditions, such as earthquake and tornado loadings as well as normal operating loading conditions. Seismic analysis of the ISFSI pad requires considering interaction between the pad and the supporting soil. Various cask loading configurations on the pad also have to be considered. Due to the lack of specific guidelines, many ISFSI pad designs show overly conservative reinforcement. This study provides guidelines and procedure for the design of the ISFSI pad that are typically used in the nuclear industry. It is considered that the guidelines and practices described in this study help design engineers understand general guidance provided in the NRC Standard Review Plans.

Abstract A free standing, slender body may experience rocking motion followed by overturning when it is subject to strong seismic motions. When the free body is submerged in water, it will also be subject to lateral forces acting along the side of the free body as a result of water sloshing. This highly non-linear situation is of particular interest to engineers in the nuclear industry in need to assess the stability of transfer casks containing spent fuel and submerged in a confined pit or pool. In this work, a three-dimensional finite element dynamic transient model of a free standing cask is developed and analyzed using ANSYS. Both dry and submerged conditions are considered. Cask to floor friction, buoyancy force, and sloshing are accounted for in the assessment. The model is validated against well-accepted contributions on sloshing and rocking provided by G.W. Housner.

Abstract History of earthquake’s damages have illustrated the high vulnerability and risks associated with failure of water transfer and distribution systems. Adequate mitigation plans to reduce such seismic risks are required for sustainable development. The first step in developing a mitigation plan is prioritizing the limited available budget to address the most critical mitigation measures. This paper presents an optimization model that can be utilized for financial resource allocation towards earthquake risk mitigation measures for water pipelines. It presents a framework that can be used by decision-makers (authorities, stockholders, owners and contractors) to structure budget allocation strategy for seismic risk mitigation measures such as repair, retrofit, and/or replacement of steel and concrete pipelines. A stochastic model is presented to establish optimal mitigation measures based on minimizing repair and retrofit costs, post-earthquake replacement costs, and especially earthquake-induced large losses. To consider the earthquake induced loss on pipelines, the indirect loss due to water shortage and business interruption in the industries which needs water is also considered. The model is applied to a pilot area to demonstrate the practical application aspects of the proposed model. Pipeline exposure database, built environment occupancy type, pipeline vulnerability functions, and regional seismic hazard characteristics are used to calculate a probabilistic seismic risk for the pilot area. The Global Earthquake Model’s (GEM) OpenQuake software is used to run various seismic risk analysis. Event-based seismic hazard and risk analyses are used to develop the hazard curves and maps in terms of peak ground velocity (PGV) for the study area. The results of this study show the variation of seismic losses and mitigation costs for pipelines located within the study area based on their location and the types of repair. Performing seismic risk analysis analyses using the proposed model provides a valuable tool for determining the risk associated with a network of pipelines in a region, and the costs of repair based on acceptable risk level. It can be used for decision making and to establish type and budgets for most critical repairs for a specific region.

Abstract The use of stainless steel (SS) tubes in boiling applications must consider the potential risk of chloride stress corrosion cracking (SCC). Most steam generating tube bundles are carbon steel or low alloy steels, but occasionally, higher alloys are needed for the process side corrosion resistance. The use of SS tubes for these cases has had both successes and failures. SS has performed very well in other water and steam services, such as condensing, steam superheating, and boiler feed water (BFW) preheating applications, but for steam generating (i.e. boiling services) the experience has been mixed. Similar failures have also occurred in various process services which are being heated and contain water. The boiling of the water can lead to SCC. Some of the variables that can affect the risk of SCC for SS tube bundles in boiling services include: chloride concentration, tube wall temperature, exchanger design (i.e. kettle, thermosiphons, etc.), vertical vs. horizontal tubes, full vaporization vs. partial vaporization, recirculation rate, and BFW blow down rate. If SS materials are being considered, the risk of SCC can be determined by analyzing these variables as described in this paper. Where the risk of SCC cannot be avoided, an alternate, resistant tube material should be selected. The material options for various services are presented herein.

Abstract The Shield Tank and End Shield Cooling System in the CANDU reactor contains a large volume of light water surrounding the Calandria and circulates water to remove heat that arises from the reactor core and Moderator. In a beyond design basis event that results in a severe event, progression in the absence of mitigating cooling actions could result in a large heat load being transferred to the water inside the shield tank from the calandria wall causing shield tank failure due to over pressurization. Following the 2011 events at Fukushima Daiichi Nuclear Power Plant, the adequacy of system pressure relief was assessed against severe events. Emergency mitigating equipment tie-ins for water make-up will likely limit the core damage state and prevent the need to protect the shield tank. However, Shield Tank Overpressure Protection (STOP) has been installed against severe event conditions pursuant to the CANDU defense-in-depth safety philosophy. A larger open vent line has been installed at some CANDU units on the top of the shield tank outside containment. This design routes the vent piping high enough to preclude any venting under any operational configuration and discharges back into the containment through an existing spare penetration. Vent piping is designed as Nuclear Class 2 in accordance with ASME BPVC Section III. Assessment of stresses in the modification piping was also completed for BDBEs including for a lower probability seismic event, steam venting and corresponding higher pressure and temperature conditions.

Abstract Continuous development of oilfields fosters a growing need for the simplification of oilfield surface process systems (SPSs) to reduce operating and management costs. Wells, testing stations, transferring stations, and central processing facilities are the main facilities in an SPS; pipelines are used to connect these stations. In this system, production radius (PR) is an important index to determine which transferring station can a testing station be linked to. Different simplification plans will lead to different operating and management costs in the following production period. Therefore, the simplification plan should be carefully designed to minimize cost and facilitate management. This paper proposes an optimization method for the simplification of SPSs in oilfields. First, an evaluation model is developed based on fuzzy analytical hierarchy process (FAHP) to select the transferring stations that need to be decommissioned. Second, hydraulic and thermal calculations are performed to get the data for the calculation of PRs. Third, the PRs, including oil gathering radius, water flooding radius, and hot water washing radius are computed to determine the linkage between the transferring stations and the testing stations. Finally, a construction plan is obtained for new pipelines of the testing stations. A case study is conducted to verify the effectiveness of this method. The results show that this method is suitable for the simplification of SPSs in oilfields.

Abstract Quenched and tempered high strength steel 07MnMoVR provides an excellent combination of strength and toughness potentially providing significant cost savings in petrochemical industry. Exposure to fire will subject steel to thermal conditions that may alter the material’s microstructure and properties. The extent of the fire damage and the potential reusability of the components can be evaluated by fitness-for-service (FFS) assessment after a fire event. According to API 579-1/ASME FFS-1, metallurgical investigation and mechanical testing are the chief means for the assessment of fire damage. This paper presents the details of an experimental investigation on the post-fire metallographic structure and hardness of 07MnMoVR steel. Metallographic analyses and hardness testing were performed on coupons exposed to elevated temperatures varying from 550°C to 850°C for half an hour to 8 hours and then naturally cooled in air or cooled by water. The results show that the microstructure of as-received 07MnMoVR steel consisted of tempered sorbite and bainite. With increasing heat exposure temperature, bainite disappeared gradually. The recovery and recrystallization of ferrite began to occur after heat exposure at 650°C for 5hrs. When the heat exposure temperature exceeded 750°C, the effects of duration time and cooling rate on microstructure were both significant. A linear correlation is indicated by fitting the ultimate tensile strength and hardness. Due to the drastically degradation of impact toughness of 07MnMoVR steel after heat exposure exceeded 650°C, it is suggested that the removal and testing of material samples shall be utilized to evaluate the fire damage of components, besides replication or in-situ field metallography and hardness testing. This study can provide basis data and guidelines for the fitness-for-service assessment of 07MnMoVR steel suffered from a fire event.

Abstract The importance of fluid-elastic forces in tube bundle vibrations can hardly be over-emphasized, in view of their damaging potential. In the last decades, advanced models for representing fluid-elastic coupling have therefore been developed by the community of the domain. Those models are nowadays embedded in the methodologies that are used on a regular basis by both steam generators providers and operators, in order to prevent the risk of a tube failure with adequate safety margins. From an R&D point of view however, the need still remains for more advanced models of fluid-elastic coupling, in order to fully decipher the physics underlying the observed phenomena. As a consequence, new experimental flow-coupling coefficients are also required to specifically feed and validate those more sophisticated models. Recent experiments performed at CEA-Saclay suggest that the fluid stiffness and damping coefficients depend on further dimensionless parameters beyond the reduced velocity. In this work, the problem of data reduction is first revisited, in the light of dimensional analysis. For single-phase flows, it is underlined that the flow-coupling coefficients depend at least on two dimensionless parameters, namely the Reynolds number Re and the Stokes number Sk . Therefore, reducing the experimental data in terms of the compound dimensionless quantity V r = Re/Sk necessarily leads to impoverish results, hence the data dispersion. In a second step, experimental data are presented using the dimensionless numbers Re and Sk . We report experiments, for a 3 × 5 square tube bundle subjected to water transverse flow. The bundle is rigid, except for the central tube which is mounted on a flexible suspension allowing for translation motions in the lift direction. The evolutions of the flow-coupling coefficients with the flow velocity are determined using two different experimental procedures: (1) In the direct method, an harmonic motion of increasing frequency is imposed to the tube. (2) In the indirect method, the coefficients are obtained from the modal response of the tube (frequency, damping). The coefficient identification was performed well beyond the system instability boundary, by using active control, allowing an exploration of a significant range of flow velocity. For a given Sk , the results show that: (a) at low Re , the flow-coupling coefficients are close to zero; (b) at intermediate Re , the flow stabilizes the tube; (c) at high Re , the flow destabilizes the tube, leading to a damping-controlled instability at a critical Re . Reducing the data in terms of Re and Sk clarifies the various experimental “branches”, which are mixed when using V r . The two identification techniques lead to reasonably compatible fluid-elastic coefficients.

Abstract The in-plane (stream-wise) fluid-elastic instability of triangular tube arrays caused tube-to-tube wear indications as observed in the U-bend regions of tube the bundles of San Onofre Unit-3 steam generators. To understand the in-plane FEI characteristics, stability analysis using unsteady fluid forces is quite helpful. However, taking measurements of unsteady fluid forces in the in-plane direction is quite challenging as the fluid coupling forces of the surrounding tubes must be measured simultaneously for the in-plane stability analysis. In particular, taking measurements under a high pressure and high temperature steam-water conditions, similar to an SG operating condition, is extremely difficult. Recently we have been able to successfully measure unsteady fluid force using a specially designed test equipment. In the meantime, random excitation forces acting on heat transfer tubes were also measured. We calculated the stability boundary of the in-plane FEI using the unsteady fluid force measurement with this special equipment.

Abstract We consider a hanging cantilevered pipe conveying water within a water-filled container; the upper portion of the pipe is surrounded by a rigid cylindrical tube of larger diameter, forming an annular fluid-filled region around the pipe. Two flow configurations are investigated : (a) water enters the pipe at its clamped end and flows downwards, discharging at its free end into the container; the fluid exits the container by flowing upwards in the annulus and out; (b) the reverse flow arrangement: water enters the system at the upper end of the annulus and exits by flowing upwards in the pipe. The dynamics of the system is studied theoretically and experimentally for both configurations. The analytical models utilized are outlined and the experiments are described. Theory and experiment find that the system loses stability at sufficiently high flow velocity by flutter or static divergence.