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Computational occupant modelling has an effective role to play in investigating road safety. Realistic representation of occupants is very important to make investigations in virtual environment. Pregnant occupant modelling can help investigating safety for unborn occupants (fetuses) however, existing pregnant occupant models are not very realistic. Most do not anthropometrically represent pregnant women and do not include a fetus model. ‘Expecting’, a computational pregnant occupant model, developed with a view to simulate the dynamic response to crash impacts is briefly explained in this paper. The model is validated through rigid bar impacts and belt loading tests and used to simulate a wide range of impacts. ‘Expecting’, possess the anthropometric properties of a 5th percentile female at around the 38th week of pregnancy. The model is complete with a finite element uterus and a realistic multibody fetus which is a novel feature in models of this kind. In this paper, the effect of further developments to ‘Expecting’, by incorporating a finite element fetus head model is investigated. Further detailed anatomic geometry is used to generate deformable fetus head model. The model is used to simulate a range of frontal impacts with seatbelt and airbag, as well as no restraint cases. The strains developed in the utero-placental interface are used as the main criteria for fetus safety. The effect of incorporating a finite element fetus head in the pregnant occupant model is discussed.

Conventional pneumatic systems present several advantages over competitive technologies, as they do not exhibit significant heat or magnetic fields and present high force to volume ratios. Nevertheless, they are very difficult to control due to their inherent nonlinearities like friction forces or air compressibility. As a consequence, their use is limited to simple motion tasks. This paper uses a nonlinear control law, previously developed by the authors, to control a servopneumatic system in the micro meter range. Preliminary experimental results show that the control law leads to very accurate motion control in micro meter positioning tasks in any position of the piston stroke. The paper also shows, through experimental results, that the control law is robust to parametric variations (change of payloads) and to external constant disturbances (gravity force).Copyright © 2014 by ASME

This paper provides a part of series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”. Paper is focused on the seismic evaluation method of the multiply supported systems, as the one of the design methodology adopted in the equipment and piping system of the seismic isolated nuclear power plant in Japan. Many of the piping systems are multiply supported over different floor levels in the reactor building, and some of the piping systems are carried over to the adjacent building. Although Independent Support Motion (ISM) method has been widely applied in such a multiply supported seismic design of nuclear power plant, it is noted that the shortcoming of ignoring correlations between each excitations is frequently misleaded to the over-estimated design. Application of Cross-oscillator, Cross-Floor response Spectrum (CCFS) method, proposed by A. Asfura and A. D. Kiureghian[1] shall be considered to be the excellent solution to the problems as mentioned above. So, we have introduced the algorithm of CCFS method to the FEM program. The seismic responses of the benchmark model of multiply supported piping system are evaluated under various combination methods of ISM and CCFS, comparing to the exact solutions of Time History analysis method. As the result, it is demonstrated that the CCFS method shows excellent agreement to the responses of Time History analysis, and the CCFS method shall be one of the effective and practical design method of multiply supported systems.

High pressure flange connections for LDPE plants are designed using lens ring gaskets and bolted flange connections. The sealing principle is to achieve a high contact pressure between the spherical lens shape and the conical tube end. This contact pressure must exceed the fluid pressure by a safety margin under all operating conditions. Bending moments acting on the flange connection will create an uneven distribution of contact pressure over the lens ring surface. In this paper a procedure is shown to analyse the influence of the bending moment under consideration of internal pressure on a flange connection using 3-dimensional finite element models. The variations in bolt stress and distribution of contact pressures due to the bending moment are of specific interest. A first approach will be provided to derive some simplified design rules regarding leakage assessment of such flange connections.

Because “safety” is still the key point in the Nuclear industry, more than in any others, the use of new materials or new manufacturing methods is really challenging. Among many characterizations required for a new steel grade or a new manufacturing process to be accepted and then introduced in a Nuclear design code, the fatigue properties must be determined with great care.Nowadays, the consideration of the Pressurized Water Reactor (PWR) primary water environment effect on the Low Cycle Fatigue (LCF) behavior of Austenitic Stainless Steels (ASS’s) is an important issue for both Nuclear Power Plants (NPP) lifetime extensions and new builds as described in the NUREG/CR-6909 [1], [2].This paper aims to present the LCF behaviors in air and in PWR water at 300°C of type 304L and 316L ASS’s manufactured by Powder Metallurgy coupled with Hot Isostatic Pressing process (PM/HIP) and to compare them with those observed on usual ASS nuclear grade products [3]-[6].As already introduced in our previous paper [7] dedicated only to the PM/HIP 304L steel fatigue behavior, it appears that the microstructures, mechanical properties and LCF behaviors in air and in PWR water of both type 304L and 316L steels manufactured by PM/HIP process are better or at least similar to those observed on wrought ASS’s.Copyright © 2015 by ASME

The Pacific Northwest National Laboratory (PNNL) is conducting studies for the U.S. Nuclear Regulatory Commission (NRC) to assess the capability, effectiveness, and reliability of ultrasonic testing (UT) as a replacement method for radiographic testing (RT) for volumetric examination of nuclear power plant (NPP) components. This particular study focused on evaluating the use of phased-array (PA) UT on carbon steel plate welds. Welding fabrication flaws included a combination of planar and volumetric types; for example, incomplete fusion, lack of penetration, cracks, porosity, and slag inclusions. The examinations were conducted using PAUT techniques applied primarily for detection and flaw type characterization. This paper will discuss the results of using PAUT in lieu of RT for detection and classification of fabrication flaws in carbon steel plate welds.

Our experimental research on fatigue performance of stainless steels and transferability of laboratory data to plant operational conditions focuses in niobium stabilized stainless steel (1.4550, X6CrNiNb1810mod) taken from a pipe manufactured as primary piping for a German NPP Good fatigue performance both in air and in PWR water was reported in previous PVP papers. The NUREG/CR-6909 report proposes Fen factors based on stroke controlled experiments in hot water for non-stabilized steels. Since PVP2013-97500 we have new data in 200°C PWR water to compare with predictions by NUREG/CR-6909. Our strain controlled tests in 325°C and 200°C PWR water give longer lives resp. smaller Fen factors. For the slowest tested strain rate 4·10−6 in 325°C water the prediction according to NUREG/CR-6909 goes just below the current ASME design curve, but our results remains well above. Including also the relevant design temperature effect, our result Fen = 4 is well below the predicted Fen = 14,5. The gap is smaller for higher strain rates and low Fen values. Simplified simulations of fatigue transients combined with normal operation indicated that relevant loading patterns as hold-time effects may result to notably longer lives than in standard laboratory tests. A concern was raised on transferability of data to thermal transients separated with months of normal operation. Cyclic strain (transients) followed by hot holds (normal operation) lead to time and temperature dependent hardening with reduction in cyclic plastic strain and fatigue usage, i.e. extension of life. This paper reports new data, challenges met and our progress towards developing realistic design factors for effects both reducing and extending fatigue endurance in nuclear power plant operational conditions.

Weld residual stress (WRS) distributions are an important input into fracture mechanics evaluations necessary to determine the residual lives of dissimilar metal welds (DMWs). Since the DMW geometry and the presence or absence, size, and location of weld repairs is nozzle specific, finite element WRS analysis is often used to predict through-wall weld residual stress distributions. It is important to note that despite small differences in plant specific geometry or weld location specific weld repair geometry there are substantial similarities between the configurations that have been evaluated in the numerous weld specific finite element WRS analyses documented in the literature. Important insight can be gained from parametric studies of simplified geometries in order to understand the significance of different parameters on the resulting WRS distributions. The results of such studies can allow engineers to focus resources on refining accuracy of critical inputs and to support simplified model development suitable for incorporation into design and fitness for service codes. This paper documents the results of various studies performed to validate the ability to use a simplified pipe-to-pipe model for simulating relative effects on through-wall WRS distributions of pipe and weld repair geometry, investigate the effect of pipe mean radius to wall thickness ratio, weld repair depth (ID and OD), and weld repair sequence. Fifteen cases are analyzed. The dimensions selected for each case span a range of wall thickness, Rm/t and depth of repair values representative of typical Boiling Water Reactor (BWR) nozzle DMWs. The results are used as input into a simplified WRS model presented in a separate paper [17].Copyright © 2014 by ASME

As an initial step in our effort to investigate the interaction of a turbulent channel flow with a compliant wall, this paper focuses on the measurement techniques. Two noninvasive optical techniques, namely tomographic PIV (TPIV) and Mach-Zehnder Interferometry (MZI), are integrated to simultaneously measure the time-resolved, wall-normal deformation of the compliant transparent wall and the 3D velocity field of a turbulent channel flow above it. The two systems utilize the same laser, but different cameras. The paper provides a description of the optical setup, detailed information about calibration of the MZI system, as well as sample combined 3D velocity distributions and wall deformations. The measured wall deformation can be decomposed into low frequency structure modes, and higher frequency features that appear to advect with the flow.Copyright © 2014 by ASME

In order to calculate the crack propagation in complicated-shaped locations in components such as weld in penetration structures of reactor pressure vessel of nuclear power plants, an automatic 3D finite element crack propagation system (CRACK-FEM) has been developed by the Nuclear Regulation Authority (NRA, Japan). To confirm the accuracy and effectiveness of this analysis system, a verification analysis was performed. The program used for comparison is PipeFracCAE developed by Engineering Mechanics Corporation of Columbus, which has been used for many crack propagation analyses in various applications. In this paper, the axial crack propagation analysis for primary water stress corrosion cracking (PWSCC) in a steam generator inlet nozzle of a pressurized water reactor (PWR) plant is presented. The results demonstrate that the two codes are in good agreement. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Guidance for weld residual stress (WRS) in flaw assessments of in-service equipment is given in Annex E of the Fitness-for-Service (FFS) standard, API 579-1/ASME FFS-1. Recent progress in the simulation of weld residual stress, and in its verification and validation has provided the opportunity to both simplify and supplement Annex E. Specifically, a tiered approach is recommended with the baseline guidance simplified and harmonized with other existing guidance, while more complicated cases are recommended to be handled with an explicitly described simulation method. The simulation method is the subject of related papers, while this paper focuses on comparison and evaluation of existing guidance for the common cases of cylinder girth and seam welds. A complete draft of recommended guidance has been completed, and intended for publication in the 2014 revision of API 579-1/ASME FFS-1. A supporting basis document which details the underlying methodologies for recommendations has also been completed.Copyright © 2014 by ASME

A previous PVP paper [1] identified suggested improvements to be made to ASME Code, Section XI, Nonmandatory Appendix G, “Fracture Toughness Criteria for Protection Against Failure” [2]. That paper also identified that the current version of Appendix G does not have any provisions for when the calculated operating stress (pseudo stress) exceeds the material yield strength. The treatment of stresses exceeding yield was included in earlier versions of Appendix G, but it was removed via Code Action ISI-94-40 in 1995. The specific reasons for removal of these provisions were not documented. In some Appendix G postulated flaw evaluations for pressure-temperature (P-T) limits, the calculated total linear-elastic (or pseudo) stress (i.e., including the primary stress due to pressure loading and thermal stress) may exceed the material yield stress. The ASME Section XI Working Group on Operating Plant Criteria (WGOPC) decided that this provision needed to be more fully considered, with appropriate benchmarking and possible adjustments to Appendix G made consistent with the current state of knowledge in elastic-plastic fracture mechanics (EPFM) methods. This is appropriate since the state of knowledge in EPFM has significantly advanced since the time the technical basis for Appendix G was established, as documented in Welding Research Council (WRC) Bulletin No. WRC-175, which was published in 1972. Furthermore, EPFM provides an improved method for evaluating the effects of high stresses. This paper describes the results of preliminary investigations of stresses exceeding the material yield stress in fracture toughness assessments associated with Appendix G. Also included in the technical evaluations presented are the temperature conditions for which upper shelf conditions are present and where EPFM methods are applicable.

Booster, Primary and Hyper compressors are the pieces of machinery used to reach the extremely high pressures needed for the reaction in Low-Density Polyethylene plants. Wide operating ranges are necessary due to the different products and conditions required by the process to obtain both homo and co-polymers required by the market. A suitable capacity control system has to be studied during the design phase to optimize the operation particularly for Booster and Primary compressors, that are supplying the required flow to the Hyper, then compressed to the reactor. In the latter the traditional capacity control cannot be used because of the very high pressures involved and for start-up and various preparation activities the by-pass along the reactor is the only flow regulation possibility. On the contrary, in Booster and Primary there is a wider choice of CCS, from “clearance pockets” to “variable speed drive”, from step to step-less. This paper gives a brief overview of these systems, explains their pros and cons and examines the impact they have on valve behaviour explaining the CCS-valve interactions. In particular it focuses on the new generation of electro-mechanical step-less control systems.

Weld residual stress (WRS) is known to be an important driver for stress corrosion cracking. Therefore, in probabilistic Leak-Before-Break calculations for safety-related nuclear piping systems, it is necessary to offer best-estimate WRS inputs along with an informed characterization of the associated uncertainty. This paper discusses the finite element (FE) analysis work performed to develop the WRS inputs for probabilistic models used for dissimilar metal (DM) welds at pressurized water reactors (PWRs). Three different weld geometry cases were considered representing typical nuclear piping system DM welds. In accounting for modeling uncertainty, the FE results from three independent analysts for the various weld geometry cases were compared and evaluated. This paper describes the modeling strategies used to develop the best-estimate inputs and the uncertainty distributions, including approaches to eliminate unnecessary sources of uncertainty, such as inconsistent post processing.

This paper provides a historical perspective on the need for, and development of, buried and underground piping tanks programs at nuclear power plants. Nuclear power plant license renewal activities, Nuclear Regulatory Commission Buried Piping Action Plan, and the rationale for addressing the issue of buried pipe through an industry initiative as opposed to regulation are discussed. The paper also addresses current NRC activities including the results of Nuclear Regulatory Commission inspections of buried piping programs at nuclear power plants as well as Nuclear Regulatory Commission involvement in industry and standards development organizations. Finally, the paper outlines the Nuclear Regulatory Commission’s future plans concerning the issue of buried piping at US nuclear power plants.

The 45-degree laterals are widely used in pressure vessel nozzles and piping branch connections. Though the pressure design is always important for the 45-degree laterals, it is not a simple work because it has severe stress concentrations, it is difficult to weld and inspect, and there are some discrepancy between a conventional design and design by linear and nonlinear finite element analysis. In previous papers, authors studied the characteristics of both 90 degree tee and 45 degree laterals using an inelastic finite element analysis based on simplified shell element models and proposed Collapse Strength Reduction Factor (CSRF) based on an inelastic analysis were compared. In this paper, results of the burst test of 45-degree lateral and 90 degree intersection were shown. The fracture surface of 45-degree lateral was different from that of 90-degree intersection. These experimental results are compared with the inelastic finite element analysis results focusing on the local stress and strain behaviors. It was found that the magnitude of the local strain affected the burst pressure. Consideration should be given on the local failure due to excessive plastic strain under high stress triaxiality for the design of the 45-degree lateral by inelastic analysis.Copyright © 2014 by ASME

Part 1 of this paper was presented by Mr. Chris Hinnant at the 2008 PVP Conference in Chicago. Since 2008 additional fatigue test data has become available to the Paulin Research Group (PRG) which includes tests on unreinforced fabricated tees intended to support the fatigue curve approach established in the 2008 paper, and twenty-five additional straight pipe cantilever fatigue tests on carbon, stainless, duplex, CuNi, P91, X42 and X65 materials. These more recent experimental results confirm the fatigue slope recommended for cantilevers in the 2008 paper, but suggest that the original Markl slope may be more suited for some configurations of branch connections. Comparisons of the B31 Code allowable with failure data, tests, reviews of ratcheting behavior in pipe systems, and crack growth monitoring help draw conclusions about design equations in the B31 Codes and fatigue test procedures.

Based on the best-estimate program RELAP5/MOD4.0, a full-scope thermal-hydraulic model with reference to CPR1000 nuclear power plant is established in this paper, which includes the thermal-hydraulic systems of conventional island as well as the primary nuclear island which has already been researched in traditional safety analysis. Therefore, this paper mainly details the numerical model of the turbine and other parts of the conventional island thermodynamic system. A comparison between the calculated results in steady-state and the actual data of reactor demonstrates a fine consistency, thus verifying the accuracy and reliability of the model. In addition, the steam parameter changes are numerically simulated during the steam turbine’s off-design operating condition such as back pressure variation and the variation trends are the same as the actual situation of nuclear power plants.Copyright © 2014 by ASME

In line with the cooperative project “Non-invasive Condition Monitoring of Nuclear Reactors for the Detection of Level Change and Deformation of the Core” between the Technical University Dresden and the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of Zittau/Goerlitz University of Applied Sciences, a measuring system for the core state diagnosis during a core melt accident in the reactor pressure vessel of a pressurized water reactor is going to be developed. The operational principle of this system is based on the non-invasive measurement of continuously changing gamma radiations (caused by the shifting of melted materials and fission products) outside of the reactor pressure vessel by means of several gamma radiation sensors. The sensors are arranged over the height of the core and the lower head. By using computer based and real-time capable methods for evaluation of the measured gamma radiations conclusions about the core state can subsequently be drawn. This paper includes a description of a concept as well as several methods for the core state diagnosis during a core melt accident. The main part is the analysis of the methods for the core state diagnosis based on different evaluation criteria.Copyright © 2014 by ASME

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

Different deaerators have different structures, and the intensities of deaerating are different. But the deaerator models without considering the structure couldn’t show these aspects. In order to ensure the accuracy of simulation results, concrete structure of the deaerator should be taken into consideration. This paper carried out a mathematical model of horizontal type deaerator with constant nozzles and trays in nuclear power plants. It was built based on the structure of the segmentation deaerators and the process of heat transferring. This paper calculated the vapors condensation rate, it was based on heat transfer coefficient of direct contact condensation, and the heat transfer area of water film when working conditions were corresponding to the empirical formulas. While the working conditions were beyond the limits of empirical formulas, this paper would build models by conservation of energy, so the model could work under any working conditions. These models were installed in the simulation system of Qinshan Phase II., and were tested under variable power and accidental conditions. The testing results show that the models could fit for different steady working conditions. Compared with the outputs and the actual operating data, the error was small. Under the conditions of variable power and turbine tripping, changing of the parameters have the same trend curve with the actual operating data. Because of taking the concrete structure into consideration, these models can be performed better of the specific characteristics of the horizontal type deaerators with constant nozzles and trays in nuclear power plants, therefore the results are more accurate than the models without considering the structure.

A Field Programmable Gate Array (FPGA) is a type of integrated circuit (IC), which is programmed after it is manufactured. FPGAs are referred to as a form of programmable hardware, as there is typically no software or operating system running on the FPGA itself. A significant amount of design work has been performed regarding the application of FPGAs in the nuclear field in recent years, with much of that work centered around safety related Instrumentation and Control (I&C) systems and safety systems. These new FPGA based systems are considered to be viable alternatives to replace many old I&C systems that are commonly used in Nuclear Power Plants (NPPs). Many of these older analog and digital systems are obsolete, and it has become increasingly difficult to maintain and repair them. FPGAs possess certain advantages over traditional analog circuits, PLCs and microprocessors, when considering nuclear I&C and safety system applications. This paper describes how FPGA technology has been used to construct a lab-scale implementation of a Post-Accident Monitoring System (PAMS), for a Westinghouse AP1000 Nuclear Power Plant, using a National Instruments “cRIO” chassis and I/O modules. This system will perform the major functions of the existing PAMS, including monitoring the vital values such as temperature, water level, pressure, flow rate, radiation levels and neutron flux in the event of a serious reactor accident. These values are required in standards such as United States Nuclear Regulatory Commission (NRC), Canadian Nuclear Safety Commission (CNSC), International Electrotechnical Commission (IEC), and Institute of Electrical and Electronics Engineers (IEEE). All of the input signals are read and processed using the FPGA, which includes alarms if the values go beyond the specified range, or if the values change rapidly. The values were then output to the computer through the FPGA interface to provide information to the operator, as well as being sent through analog and digital output modules for further processing. The system was tested using both simulated and real inputs from sensors. Furthermore, the reliability of the new system has also been analyzed, using the Dynamic Flowgraph Methodology (DFM). DFM has been successfully applied in both the nuclear and aerospace fields, and has been described as one of the best methodologies for modeling software/hardware interactions, by the scientific literature as well as in NRC reports. DFM was applied to fine-tune the design parameters by determining the potential causes of faults in the design, as well as to highlight the effectiveness of DFM in nuclear and in FPGA applications.

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical-energy generation. The largest group of operating Nuclear Power Plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) have gross thermal efficiencies ranging from 30% and up to 36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8 MPa / 257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fuel – natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., SuperCritical Water-cooled Reactors (SCWRs) have to be designed. This path of the thermal-efficiency increasing is considered as a conventional way through which coal-fired power plants gone more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel Pressure-Channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulic code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the CFD Fluent code has been used for better understanding of specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical-water loop and developing passive-safety systems.

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.Copyright © 2014 by ASME

Competition in the electricity market forces producers to achieve — in compliance with safety — efficiency of production as high as possible. This efficiency and heat rate is an important indicator of both the condition of the power plant equipment and the quality of power plant operation. To cope with these challenges, powerful methods are process data reconciliation, statistical data processing of large data sets and process simulation. These functions and methods can be used to obtain useful information about process quality and equipment and sensor health. The paper discusses practical experience from six years of using a thermal performance monitoring and optimization system in the Dukovany nuclear power plant. The system is integrated into the overall nuclear power plant process information system and data warehouse. The system provides information in near real time. The major benefit of the system lies in a deep view into equipment behaviour and process which ensures timely detection and identification of functionality degradation of process and equipment or sensor faults. The system also helps to find and use margins of equipment operation and the overall thermal cycle. Selected practical examples are used to demonstrate specific benefits of the system for operation and maintenance of the Dukovany nuclear power plant. There are examples of equipment fault detection and sensor degradation detection. The optimization function is explained with an example of cooling circuit optimization aimed to increase the delivery of electrical power into the grid. A detailed description of behaviour of the main components can be used for their performance evaluation and their repair planning. The benefit of more accurate determination of parameter values is reflected in more accurate determination of reactor thermal output. The conclusion of the paper provides an overall evaluation of system benefits for operation and maintenance of a nuclear power plant.

There is a huge and rapid growing demand for energy in China. The developing of new energy, including nuclear energy, is an important way to solve China’s energy and environment problems. Currently, Nuclear power is playing a more and more important role in China, especially in the coastal areas where the economy is developing rapidly. The main nuclear reactor equipment is the key guarantee to nuclear safety, such as the reactor pressure vessel. With the continuous development of science and technology, the reactor pressure vessel uses more and more of advanced design methods, high performance materials and industrial-proven technologies. As an airtight container who is directly exposed to the great operating pressure of reactor, the reactor pressure vessel plays a crucial role to the safety of reactor. In the paper, by retrieving Chinese Intellectual Property Office patent libraries, 12 Chinese innovative patents for the reactor pressure vessel are introduced, containing four kinds of main technologies: closure, body, nozzle safe end and other technologies. The patent results could lead the corresponding advice to the design of the equipment effectively, to the system arrangement reasonably, and further to the reference for designing a safe nuclear power plants. The paper is to exchange the latest technological progress of reactor pressure vessel, in order to promote technological innovation in the nuclear power field.Copyright © 2014 by ASME

This paper briefly introduces the safety instrumentation and control (I&C) system (Teleperm® XS) designed for the nuclear power plant in Mochovce units 3 and 4 (Slovak Republic). The overall I&C architecture of the concerned nuclear power plant is shortly introduced as well. An overview is given on the different test phases for the hardware and software I&C modules. The integrated I&C test concept and its implementation is presented as well as the description of the integrated test phase in the test bay in Erlangen (Germany). After a successful completion of the integrated test phase for unit 3, the Factory Acceptance Test (FAT) and the erection phase have been started for the concerned I&C safety systems (e.g., the Reactor Protection System and the Reactor Power Limitation System). This paper will also present the significant advantages and specifics of performing the concerned I&C tests in the aforementioned test bay.

Taking into account the expected increase in global energy demands and increasing climate change issues there is a pressing need to develop new environmentally sustainable energy systems. Nuclear energy will play a big part in being part of the energy mix since it offers a relatively clean, safe and reliable source of energy. However, opportunities for building new generation nuclear systems will depend on their economic and safety attractiveness as well as flexibility in design to adapt to different countries and situational needs. Keeping these objectives in mind, a framework for international cooperation was set forth in a charter of Generation IV International Forum (GIF, 2002). The main design goals for the Generation IV nuclear system concept were to improve economic gains, enhance safety, extend sustainability, and strengthen proliferation resistance. To achieve high thermal efficiencies of up to 45–50%, the use of SuperCritical Fluids (SCFs) as working fluids in heat transfer cycles is proposed. An important step towards development of SCF applications in novel Generation IV Nuclear Power Plant (NPP) designs and in other industries; is to understand the thermal hydraulic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. Heat transfer under SC turbulent conditions is generally very complex and is extremely sensitive to the test geometry and operational flow parameters. Detailed sets of experiments have been conducted around the world in tubes and other geometries with SCFs to study the basics of heat transfer. By variation of operational parameters and test geometries, fundamental heat transfer data sets are collected that will help in our understanding of SC heat transfer phenomena. Applications for SCFs are not limited to power industry; recent advancements have indicated the use of SCFs in a much wider range of applications due to its unique and attractive heat transfer characteristics. In this paper, proposed uses of SCFs are presented and basic concept of DHT within SCFs is analyzed.

This paper uses the China AP1000 project as an example to exhibit the application of quantitative risk management in nuclear power plant construction projects. For those lump sum contracts, one of the most significant purposes of quantitative risk management is to determine the contingency, i.e. the reserved money and time for projects. This paper studies the application of Monte Carlo simulation in determining the contingency, taking into account the distinctive features of nuclear power projects. Most nuclear power projects, especially advanced ones such as Generation III and above, meet one common obstacle in estimating key economic indicators — the absence of historical data due to its avant-garde design. As cost estimators of the coal power plant contractors may collect their data from thousands of previous cases, nuclear power plant contractors, especially in many developing countries, do not have a shared database of financial data. Some first-of-a-kind nuclear power plants have absolutely no historical data to look up. This paper aims to provide a resolution to this problem. First, the feasibility and representativeness of different probability distributions are compared based on their respective skewness and kurtosis to determine the best-suited distribution in nuclear power projects. This paper also analyzes the use of second-order Monte Carlo simulation in reducing the error caused by the biased estimation of inexperienced risk assessment engineers.

The active magnetic bearing (AMB) system is a crucial part in the helium circulator system of the 10MW high temperature gas-cooled reactor (HTR-10). Though the AMB has been widely used in industrial fields, it is still limited in the research of the dynamic behavior of AMB’s vertical arranged rotor with axial magnetic load during its drop process. This paper establishes the dynamic model of such drop process by Matlab. Meanwhile using the Hertz contact theory establishes the contact model of different configurations. Analyze the axial friction between the rotor and thrust interface of the inner ring of Auxiliary Bearing System (ABS). Besides, the numerical model is verified by the drop experiment with the axial magnetic force. Moreover, this paper analyzes the influence of the rotor’s drop rotational frequency and the axial bracing features including stiffness and damping on the dynamic behavior during vertical arranged rotor’s drop process. Moreover, the paper provides the optimal axial stiffness and damping for the ABS satisfying the experimental conditions so as to reduce the contact force. Such results provide important references to the design of the ABS with a vertical arranged rotor and its application in HTR-10 and High Temperature Reactor-Pebblebed Modules (HTR-PM).Copyright © 2014 by ASME

As an essential component in the system of the HTR-10 helium turbine generator electromagnetic bearings, auxiliary bearings provide mechanical backup protection in case of the events of magnetic bearing failure happen. When contact events happen, highly localized and transient temperatures will arise from frictional heating over the dynamically varying contact area in the very short term and dissipate through the bearing in the longer term. When excessive temperature level occurs, rapid failure may be anticipated, thus it will become a serious threat to the safety of the HTR. This paper presents a detailed analysis of thermal growths due to the mechanical rub for a rotor drop on auxiliary bearings. With the aim to numerically analyze the heat generation and temperature rise, a 1D thermal model of the ball bearing composed of heat transfer network and heat sources based on heat transfer equations is established. The Matlab codes are developed to complete the numerical analysis, and an infrared method is utilized to investigate the temperature rise at the rotor/inner race contact surface. By the comparison between simulation results and the experimental data, this paper illustrates the thermal growth during a rotor drop process, which is highly non-linear. The results reveal that the axial contact force is critical to the bearing heat generation, and the ceramic balls with superior thermal properties are recommended.Copyright © 2014 by ASME

Modern steam turbine designs are exploring wider design space to gain improved performance. At the same time, the mechanical integrity of the steam turbine needs to be maintained. This requires an in-depth understanding of the effects of each design variable on High Cycle Fatigue (HCF) of a steam turbine stage. In a previous paper, effects of both Blade Count Ratio (BCR) and Gap Ratio (GR) were studied in a steam turbine stage. This paper studies the HCF of a new rotor blade with a narrower chord, which shows higher response. Furthermore, the effect of stagger angle of the stator blade is investigated. Their effects on aero forcing, aero damping and frequency shift are also demonstrated. A one-way coupling approach is used in this study due to its high computing efficiency.Copyright © 2014 by Alstom Technologie AG

This paper provides the first step in sizing carbon brush seals for aero-engine oil bearing chambers applications. Recent developments in the aeronautic domain focus strongly on the reduction of aero-engine specific oil consumption. For instance, optimizing the civil aircraft gas turbine engine lubrication oil system is considered as one of the main targets in this research. Specifically, brush seals have shown tremendous leakage performance in sealing secondary flows compared to classic labyrinth seals over the last few decades. Therefore, an attractive idea is to extent their utilization to oil bearing chamber applications. To perform the experimental part of the study, a test rig has recently been built at ULB. This test rig will be described in this paper. A parametrical study has been performed in stationary conditions, and at very low rotation speed. A particular attention was given to the air consumption and the torque friction losses. Finally, a test simulating the effects of a rotor excursion on the brush seal performance has been made.

The present paper describes an experimental investigation of the noise emission of a full-scale Garrett (Honeywell) GTCP 36-300 APU as installed on commercial airplanes like the Airbus A320. The APU exhaust noise emissions are measured in a well-controlled environment by 25 ground-based microphones arranged in a polar arc. Comparing a hard-wall reference case with different exhaust mufflers, special attention is given to the assessment of the noise reduction obtained. The performance of a newly designed muffler is compared to the standard device for the most relevant operating states ECS (environmental control system) and MES (main engine start). The paper concludes with a short examination of possibilities to improve the overall noise reduction.Copyright © 2014 by ASME

This paper is devoted to the unbalance identification and balancing of an aircraft engine rotor running on Squeeze Film Damper (SFD) bearings that show highly nonlinear features. The high pressure (HP) rotor in a twin-spool assembly cannot be accessed under operational conditions because of the restricted space for instrumentation and temperatures that are beyond the safe operating limits of the sensors. This motivates the use of a non-invasive procedure, requiring prior knowledge of the structure. The only such method for rotating machinery with non-linear bearings reported in the literature is highly limited in its application (e.g. assumes circular centred orbits). The methodology proposed in this paper is aimed at overcoming such limitations. It uses the Receptance Harmonic Balance Method (RHBM) to generate the backward operator using vibration measurements taken from sensors installed on the engine casing. The operator is then inverted using either Least Squares Fit (LSF) or Singular Value Decomposition (SVD). The resulting solution is the equivalent unbalance distribution in prescribed unbalance planes of the HP rotor which are consequently used to balance it. This method is validated using two distinct rotordynamic systems and simulated casing vibration readings corrupted by different noise levels.Copyright © 2014 by ASME

Variable Stator Vanes (VSVs) are commonly used in multi-stage axial compressors for stage matching at part load operations and during start up. Improper VSVs settings or malfunction of the controlling actuator system can lead to compressor instabilities including rotating stall and surge. It is important to be able to predict the aerodynamic behaviour of compressors in such events to either produce tolerant designs or incorporate diagnosis and recovery systems. This paper presents a numerical study of a compressor operating near the stall boundary for a mal-scheduled VSVs case. A high-speed three-stage axial compressor with Inlet Guide Vanes (IGV) is used in the investigation because of its relative simplicity and availability of geometry and aerodynamic data. A 3D RANS viscous unsteady time-accurate flow solver was used to perform the full annulus simulation with a downstream variable nozzle to control outflow boundary conditions. The unstructured mesh contained about 25 million grid points and the simulation was performed on a high performance computing cluster for many engine rotations. Rotating stall with one single cell covering several passages in all three rotors was predicted which propagated at approximately half of the shaft speed. Full analysis of the flow features is presented in the paper.

In recent years Avio Aero and the Karlsruhe Institute of Technology (KIT) have significantly enhanced their available experimental technology level in order to consolidate and deepen the investigation of some critical phenomena in aero-engine gas turbine combustion. A key parameter in the development of a lean-burn combustor is the design of the ignition system. The increase in the amount of air that flows through the primary zone to approximately 60% of the overall air proves to be beneficial for emissions reduction. In return it causes shorter residence times in the primary zone and makes the flame kernel generation and propagation a real challenge. With the scope to enhance the fundamental understanding of the ignition process at altitude conditions, the ISCRA rig (Ignition in Subatmospheric Conditions – Rig for Altitude Relight Investigation) has been designed and manufactured at the KIT. The design of the rig allows the generation of altitude conditions at variable flow velocities and turbulence characteristics. An optical access for ignition recording by a CCD camera is also provided. The paper presents the results of a fundamental investigation at the KIT using a generic setup. It consists of a pressure atomizer with known atomization properties mounted in a test rig which allows the variation of several parameters that influence ignition, namely: air pressure; air temperature; velocity; Fuel-Air-Ratio (FAR); Sauter Mean Diameter (SMD); spark energy; turbulence intensity and length scale. Two grey spots that exist in the knowledge map of the ignition process are addressed in this paper. These are data on kerosene spray ignition at temperatures below 293 K, where data is scarce and the relationship between flame kernel generation and propagation, which has not been broadly investigated nor is well understood. The ignition probabilities of kerosene sprays as a function of ignition energy in the temperature range from 253 K to 293 K, pressure of 0.7 bar, flow velocities up to 6 m/s at different turbulence intensities and length scales are measured. Furthermore, the influence of the above mentioned parameters on the kernel generation and propagation is presented and discussed. Results show that the energy required for flame kernel propagation is higher than for flame kernel generation. This difference increases at lower temperature and higher turbulence intensity. The paper gives an insight in magnitude of this difference in energy and compares the tendencies with an earlier model. The absolute values need to and will be corrected by future in situ measurements.

A modern gas turbine engine (GTE) is a complex non-linear dynamic system with the mutual effect of gas-dynamic and thermal processes in its components. The engine development requires the precise real-time simulation of all main operating modes. One of the most complex operating modes for modeling is “cold stabilization”, which is the rotors acceleration without completely heated up the turbine elements. The dynamic heating problem is a topical practical issue. Solving the problem requires coordinating a gas-path model with heat and stress models, which is also a significant scientific problem. The phenomenon of interest is the radial clearances change during engines operation and its influence on engines static and dynamic performances. To consider the clearance change, it is necessary to synthesize the quick proceeding stress-state models (QPSSM) of a rotor and a casing for the initial temperature and dynamic heating. The unique feature of the QPSSM of GTEs is separate equation sets, which allow the heat exchange between structure elements and the gas (air) and the displacements of the turbine rotor and the casing. This ability appears as a result of determining the effect of each factor on different structural elements of the engine. The presented method significantly simplifies the model identification, which can be performed based on a precise calculation of the unsteady temperature fields of the structural elements and the variation of the radial clearance. Thus, the present paper addresses a new method to model the engine dynamics considering its heating up. The method is based on the integration of three models: the gas-path dynamics model, the clearance dynamics model and the model of the clearance effect on the efficiency. The paper also comprises the program implementation of the models. The method was tested by applying to a particular turbofan engine.

This paper presents an experimental and computational investigation of the influence of volute cross-sectional shape on the performance of a radial turbocharger turbine under pulsating conditions. Two volute configurations (denoted volute A and B) with the same area-to-radius ratio (A/R) distribution but different aspect ratios are rapid prototyped and tested with a same radial rotor. Experimental results show that the turbine with smaller aspect ratio volute (volute A), which has a squarish shape, shows consistently better cycle averaged efficiency at different loadings and frequencies, and the magnitude of improvement is influenced by the operational conditions. In consequent to experiments, a reduced order unsteady computational fluid dynamics (CFD) method was employed to investigate the mechanism of the performance discrepancies between volute A and B. Computational results show a stronger flow angle distortion in both circumferential and spanwise direction for volute B. Furthermore, compared with the volute A, the flow distortion near the shroud at the rotor inlet is evidently amplified by the volute B under pulsating conditions compared with the corresponding steady condition. Results in this paper, in general, demonstrate a direction for desired volute cross-sectional shape to be used in a turbocharger radial turbine.

This paper study the non-synchronous vibration (NSV) of a high speed multistage axial compressor using rigid blade and vibrating blade with fluid-structural interaction (FSI). The unsteady Reynolds-averaged Navier-Stokes (URANS) equations and mode based structural dynamic equations are solved. A low diffusion E-CUSP Reimann solver with a 3rd order WENO scheme for the inviscid fluxes and a 2nd order central differencing for the viscous terms are employed. A 1/7th annulus sector of IGV-rotor-stator is used with a time shifted phase lag BC at circumferential boundaries. An interpolation sliding boundary condition is used for the rotor-stator interaction. The URANS simulation for rigid blades shows that the leading edge (LE) tornado vortices, roughly above 80% rotor span, travel backwards relative to the rotor rotation and cause an excitation with the frequency agreeing with the measured NSV frequency. The predicted excitation frequency of the traveling vortices in the rigid blade simulation is a non-engine order frequency of 2603 Hz, which agrees very well with the NSV rig testing. For the FSI simulation, the results show that there exist two dominant frequencies in the spectrum of the blade vibration. The lower dominant frequency is close to the first bending mode. The higher dominant frequency close to the first torsional mode agrees very well with the measured NSV frequency. The simulation conducted in this paper appears to indicate that the NSV is excited by the traveling vortex.

Thermal barrier coatings (TBC), in combination with sophisticated cooling systems are crucial for the operation of highly efficient gas turbines. New generations of coatings will need to show increased cycling capability as a future energy mix will contain a high proportion of renewable energy which will be subject to rapid changes in supply. This will require gas turbines to be on stand-by to fill shortages in power supply with short notice. Furthermore, higher operating temperatures are sought to improve the efficiency of the engine. It is, therefore, an aim of the industry to find a coating composition or structure which will enable the operation at temperatures greater than 1250°C and with high cycling capability. Test methods are required to meet these new operating conditions to validate new coatings. The maximum temperature limit of commonly used isothermal or cyclic oxidation tests is usually the temperature at which the substrate will start to significantly oxidise. However, there is the technical need to test the ceramic top layer at elevated surface temperatures up to 1500°C while keeping the substrate ‘cool’. Such capability would allow the effects of ceramic sintering, and deposit induced damage to be assessed at the TBC surface. This only can be performed on a complete coating system, when a thermal gradient is established throughout the coating. This paper reviews a burner test facility, designed and built by Sensor Coating Systems Ltd. (SCS), which combines severe and frequent cycling with the exposure of the coating to high surface temperatures and active cooling of the substrate. Further, this test can include thermal shock by active cooling of the surface at the end of each cycle. The paper will consider different operating conditions and will review experiences in building and operating the rig, including results from thermal barrier coating tests on electron beam physical vapour deposition (EBPVD) and atmospheric plasma spray (APS) samples. Further, the rig is capable of testing optical techniques such as pyrometry and thermographic phosphor thermometry for measuring surface temperature in controlled laboratory conditions and example of this will be presented. The paper also will reflect on the ISO 13123:201 standard for this type of test.

This paper presents different numerical methods to predict the thermal load of a convection cooled gas turbine blade under realistic operating temperature conditions. The subject of the investigation is a gas turbine rotor blade which is equipped with a state-of-the-art convection cooling system. Firstly, two FEM based methods are introduced. One method, referred to as FEM1D method, uses empirical correlations from the open literature to obtain one dimensional heat transfer coefficients along one flow line inside the cooling channels while in the hot gas path a three dimensional CFD simulation is used. The second method (FEM2D) uses three dimensional CFD simulations to obtain two dimensional heat transfer coefficient distributions for both, the inner cooling channels and the hot gas path. The results from both numerical methods are compared with each other and are validated with experimental data, quantifying also their accuracy limits. In total this paper gives an evaluation of two different FEM methods to predict temperature distribution in convection cooled gas turbines. Their accuracy, numerical cost and limitations are evaluated. It turns out that the temperature profiles gained by both methods are generally in good agreement with the experiments. However, while causing higher numerical costs the more detailed FEM2D method achieves more accurate results.

This paper aims to numerically and experimentally analyze the effect of the micro-jet on the suction side (SS) of the transonic high-pressure turbine cascade, which could enable the possibility of the mass matching for turbine of the variable cycle engine (VCE) when it transforms working conditions according to flight demands. The experimental study had been conducted in a transonic and supersonic cascade wind tunnel to observe the influence of this kind of flow control technique on aerodynamic performance, operated at different exit Mach numbers. The nozzle guide vane with convergent-divergent passage was investigated in this paper. Two-dimensional steady simulations were performed to investigate the influence of the flow control method to exploit the potential of the application of this method in the design of high-pressure (HP) turbine for VCE. Results of the numerical simulations and experiments indicated that micro-jet on the SS of the turbine blade would generate a separation zone near the jet slot, which would reduce the genuine throat area of the turbine passage and thus change the mass flow rate through the passage of the turbine cascade significantly. Also, velocity at the passage throat would be reduced due to the application of SS micro-jet, which would also contribute to the reduction of the mass flow rate through the turbine passage. However, extended total pressure loss was introduced with SS micro-jet.

This paper presents a seismic analysis on steam turbine within a Nuclear Power Plant. Firstly, considering that steam turbine is a kind of large equipment with complex structure, high precision and tightly connected with other equipments and structure, this paper coupled together the steam turbine, condenser and its spring foundation as a whole structure for analysis, and created detailed finite analysis models for almost each part. Secondly, the detailed model of steam turbine, condenser and spring foundation was analyzed separately to obtain their dynamic properties (inertia, eigenvalues etc.) to verify whether the detailed finite analysis models were correct and proper. Thirdly, with seismic analysis on the whole structure with time-history method, relative displacement time history curves of some places, as well as stress response of LP (low pressure) outer and inner case, was extracted. Finally, the anti-seismic performance of steam turbine, such as rotor gaps using those relative displacement and stress response results, was analyzed and evaluated.Copyright © 2014 by ASME

Turbine blades and vanes of modern aero-engines are commonly manufactured by casting. The casting process often introduces slight geometry variations. In the endwall region this leads to inter-platform steps, gaps and leakage flows. This paper determines the underlying loss mechanisms associated with each of these geometric features and guides designers in minimizing their impact on efficiency. The paper shows that the presence of an inter-platform step causes a pair of vortical structures to be superimposed onto the blade existing secondary flow structure. These are shown to always increase loss. When manufacture variations are considered the optimal design intent blade is shown to be one where the suction side endwall is lower than the pressure side endwall. The paper shows that when leakage mass flow is introduced the presence of a step can either raise or reduce loss. A correlation which gives the optimal step height for a set leakage mass flow is presented. In the final part of the paper measured engine vane geometries are used to determine the impact of endwall geometry variation on turbine stage efficiency.

A radial inflow turbine installed in a large extraction facility in Canada has shown to be a significant contributor to the noise generated in that facility. These high sound pressure levels have led to the expenditure of millions of dollars in sound abatement measures over the last 20 years, and pose a health danger to employees. Based on shop testing and field data, it was believed that a new rotor design could dramatically reduce the noise generated. It was further believed that Computational Fluid Dynamics with Aeroacoustic analyses was capable of predicting this improvement. Such a process can allow parts to be designed and installed quickly, when time and budget does not allow for detailed factory testing. This paper describes this turbine redesign procedure and the details of the analyses performed throughout the process. This new rotor has now been installed on-site, and measurements have been made before and after the replacement. This offers a unique opportunity to validate the numerical work performed through this redesign, which is also presented in the paper.

Turbine rotor tips and casings are vulnerable to mechanical failures due to the extreme thermal loads they undergo during engine service. In addition to the heat flux variations during the engine transient operation, periodic unsteadiness occurs at every rotor passage, with amplitude dependent on the tip gap. The development of appropriate predictive tools and cooling schemes requires the precise understanding of the heat transfer mechanisms. The present paper analyses the nature of the overtip flow in transonic turbine rotors running at tight clearances and explores a methodology to determine the relevant flow parameters that model the heat transfer. Steady-state three-dimensional Reynolds-averaged Navier–Stokes (RANS) calculations were performed to simulate engine-like conditions considering two rotor tip gaps, 0.1% and 1%, of the blade span. At tight tip clearance, the adiabatic wall temperature is no longer independent of the solid thermal boundary conditions. The adiabatic wall temperature predicted with the linear Newton's cooling law was observed to rise to unphysical levels in certain regions within the rotor tip gap, resulting in unreliable convective heat transfer coefficients (HTCs). This paper investigates different approaches to estimate the relevant flow parameters that drive the heat transfer. A novel four-coefficient nonlinear cooling law is proposed to model the effects of temperature-dependent gas properties and of the heat transfer history. The four-parameter correlation provided reliable estimates of the convective heat transfer descriptors for the 1% tip clearance case, but failed to model the tip heat transfer of the 0.1% tip gap rotor. The present study allows experimentalists to retrieve information on the gap flow temperature and convective HTC based on the use of wall heat flux measurements. The use of nonlinear cooling laws is sought to improve the evaluation of the rotor heat transfer data while enhancing the understanding of tight-clearance overtip flows.

In the paper the authors will present the design and preliminary test results for a high specific power (i.e., kW/kg) fully integrated and completely oil-free gas turbine driven electric generating system that operates with commercially available heavy fuel. The oil-free, high-speed micro-turboalternator system achieves high specific power through operating speeds to 180,000 rpm and the use of compliant foil bearings, high performance compressor and turbine and a permanent magnet alternator. The high operating temperatures and speeds require that oil-free compliant foil bearings be used and that the alternator section be isolated from the turbine engine portion of the system. The selected modular design approach, including compressor and turbine aerodynamic design, system thermal management issues and the corresponding impact on rotor bearing system dynamics, will all be presented. The paper concludes with a presentation of preliminary testing results showing stable full speed operation and peak power generated. Data obtained compares well with design predictions both from a rotordynamic and with regard to the cycle efficiency and performance. Conclusions regarding the ability to scale the technology to even smaller systems will also be presented.

Within this paper, the axial inclination of the bristle pack as a new design parameter for brush seals for use in a steam turbine and other rotating equipment is discussed. It is widely known that the behavior of brush seals can be influenced by important main design parameters of the bristle pack such as, but not limited to, the bristle thickness, the lay angle or the bristle length. Furthermore, the variation of the front and back plate results in different seal characteristics [1]. Each one of these parameters also has an influence on bristle damping, the blow down capability and thus the leakage flow. In addition, under changing and transient operating conditions, the radial adaptivity, which is essential for accommodating shaft deflection, is also a very important property. For a comprehensive seal design, the wear characteristic and deterioration effects have to be considered beside the above mentioned properties. At the Technical University of Braunschweig, brush seals are experimentally investigated with above focus on different test rigs. These rigs allow a detailed sealing performance investigation including live bristle pack observations and blow down measurement using cold air as well as brush seal investigations using live steam conditions up to 50bars and 450°C and a rotating shaft with representative rotational velocity. The paper shows and discusses experimental results of different axial inclinations of the bristle pack, while testing with constant front and back plate designs. The influences on the blow down, the axial behavior of the bristle pack, the leakage flow and the bristle pack stiffness are shown. The new effect of a rotating blow down type of bristle oscillation is also shown and discussed and finally a classification of the seal behavior depending of the different axial inclination is given.Copyright © 2014 by ASME

In this paper some novel Smoothed Particle Hydrodynamics (SPH) concepts are presented towards a feasibility study into the use of SPH for some aero-engine applications, e.g. for internal oil or fuel applications. A first challenge is to develop a capability to model complex wall geometries, associated with two-phase flows in gear boxes and bearing chambers for example. A demonstration is made of how such complex (for SPH) geometries can be built together with an outline of some of the wall boundary condition concepts used, including moving walls. This is an important feature for the application of SPH to engineering. Other boundary conditions are needed such as inlets, outlets and pressure boundaries, and a proper treatment of the free surface. These are outlined in the context of the proposed application. From an SPH flow simulation viewpoint, one of the challenges is to reduce the non-physical density variations arising from boundary conditions (at wall, free surface and interface), which are responsible for non-physical pressure variations and particle dynamics. The flow regimes found in the engineering systems outlined above involve droplets, filaments and films. It is therefore important to be able to handle the merging of fluids, as it is to model their interaction with another phase, which calls for appropriate multi-fluid and surface tension models. This paper introduces SPH, outlines a number of concepts listed above and presents some preliminary results towards the modeling of the KIT bearing chamber, as described by Kurz et al. [1]. This work builds on a number of numerical modeling communications made by the Nottingham team to SPHERIC, the ERCOFTAC Special Interest Group (SIG) for SPH.