bluuebunny/crossref_metadata_2025 huggingface dataset:
In-Wall Network (Mesh) Cooling Augmentation of Gas Turbine Airfoils
Authors: Bunker Ronald S. , Bailey Jeremy C. , Lee Ching-Pang , Stevens Charles W.
Published: November 2008

The present study investigates a new class of in-wall mesh cooling techniques intended to produce significant wetted surface area heat transfer enhancement products, hAAwetted, with low to moderate pressure losses. Cooling networks (meshes) are presented using round pins and rounded diamond shaped pins with H/Dp ratios of 0.2 and S/Dp ratios of 1.5, as well as less dense rounded diamond pins of smaller diameter with H/Dp of 0.3 and S/Dp of 2.14. These geometries differ substantially from conventional pin fin arrays in which H/Dp ≥ 1 and S/Dp ratios are typically about 2.5. The objective of mesh cooling is to provide highly effective cooling inside component walls while maintaining very thin outer protective skin thickness. Special attention is paid to the combination of techniques including pin meshes, turbulators, and concavity arrays that can provide multiple design solutions for heat transfer and pressure loss goals. Average channel wetted-area heat transfer capabilities exceeding 3 times that of smooth channels have been demonstrated using actual surface area enhancements of no more than 20%. This cooling capability increase is also realized with a large decrease in channel material solidity, up to 30%, compared to conventional pin fin arrays. Friction factor enhancements in the range of 15 to 25 times that of a smooth channel turbulent flow accompany this performance. This investigation represents the first of its kind to demonstrate feasible in-wall cooling methods that may be realized in practice for investment cast turbine blades.
Natural Frequencies and Modes Shapes of Two Mistuned Bladed Discs on the Shaft
Authors: Rza˛dkowski Romuald , Drewczynski Marcin
Published: November 2008

The dynamic behaviour of a rotor consisting of two bladed discs on a solid shaft is considered. The effect of shaft flexibility on the dynamic characteristics of the bladed discs and the coupling effects between the shaft and bladed disc modes are investigated. Results show clearly the coupling effects in a bladed disc-shaft system. In this study, the global rotating mode shapes of flexible tuned and mistuned bladed disc-shaft assemblies are calculated. Rotational effects, such as centrifugal stiffening are accounted for. The gyroscopic effects are included. Calculated natural frequencies obtained from blade, bladed disc and shaft with two discs are checked to discover resonance conditions and the coupling effects. The torsional frequency of the shaft with two bladed discs is coupled with the zero nodal diameters modes of the single bladed discs. It is shown that blade mistuning strongly affects the interaction between flexible bladed discs and flexible shaft. This interaction affects the flexible bladed disc modes, and it is not restricted to the modes with zero, one and two nodal diameters.
Exposure Test Results of Lu2Si2O7 in Combustion Gas Flow at High Temperature and High Speed
Authors: Yuri Isao , Hisamatsu Tohru , Ueno Shunkichi , Ohji Tatsuki
Published: November 2008

In order to understand recession behavior and the amount of recession of Lu2Si2O7 in the combustion gas flow, sintered Lu2Si2O7 specimens were manufactured by hot pressing and exposed under various combustion gas flow conditions (T = 1300–1500 °C, P = 0.3 MPa, V = 150 m/s, PH2O = 27–69 kPa, t = 10h). After the exposure tests, etch pits, which are assumed to form due to volatilization of SiO2 in the grain boundary phase, were observed at the surface of specimen. The amount of Lu2SiO5 phase at the surface of specimen increased with the increase of gas temperature or water vapor partial pressure. A corresponding decrease in the amount of Lu2Si2O7 phase was observed. Furthermore, by using the average weight loss rate for exposure times of ten hours, the influence of gas temperature and water vapor partial pressure on weight loss rate was examined, and the amount of recession under gas turbine conditions was calculated.
Flow Characteristics and Stability Analysis of Variable-Density Rotating Flows in Compressor-Disk Cavities
Authors: Johnson Bruce V. , Lin J. D. , Daniels William A. , Paolillo Roger
Published: November 2008

Previous heat transfer experiments showed that significant differences in the flow and heat transfer characteristics can occur in models of aircraft gas-turbine, high-compressor drums. Experiments with heated disks and colder flow show large-scale instabilities that cause mixing between the cooling flow and the flow in the trapped cavities. The general result of this mixing is relatively high heat flux on the disks. Other heat transfer experiments, simulating the aircraft take-off condition with cold disks and hotter coolant, show decreased heat transfer due to the stabilizing effects of positive radial density gradients. A stability analysis for inviscid, variable-density flow was developed to quantify the effects of axial velocity, tangential velocity and density profiles in the bore region of the disk cavities on the stabilizing or destabilizing characteristics of the flow. The criteria from the stability analysis were used to evaluate the axial velocity and density profile conditions required to stabilize three tangential-velocity profiles, obtained from previous experiments and analyses. The results from the parametric study showed that for Rossby numbers, the ratio of axial velocity to disk bore velocity, less than 0.1, the flow can be stabilized with ratios of cavity density to coolant density of less than 1.1. However, for Rossby numbers greater than 1, the flow in the bore region is unlikely to be stabilized with a positive radial density gradient. For Rossby numbers between 0.1 and 1.0, the flow stability is a more complex relationship between the velocity and density profiles. Results from the analysis can be used to guide the correlation of experimental heat transfer data for design systems.
Recent Developments in Turbomachinery Modelling for Improved Balancing and Vibration Response Analysis
Authors: Chen Peter Y. P. , Feng Ningsheng , Hahn Eric J. , Hu Wenlong
Published: November 2008

Present day turbogenerator installations are statically indeterminate rotor-bearing-foundation systems utilising non-linear hydrodynamic bearings. For optimal balancing and diagnostic purposes it is important to be able to correctly predict the system vibration behaviour over the operating speed range. Essential aspects of this involve identifying the unbalance state, identifying appropriate dynamic foundation parameters, and identifying the system configuration state (relative location of the support bearings). This paper shows that provided the system response is periodic at some speeds over the operating range, and provided appropriate rotor and bearing housing motion measurements are made, it is possible in principle to satisfactorily achieve the above identifications without relying on Reynolds equation for evaluating bearing forces; and preliminary results indicate that the identifications achieved promise to be superior to identification approaches which use Reynolds equation.
Parametric Optimization of a High-Lift Turbine Vane
Authors: Arnone Andrea , Bonaiuti Duccio , Focacci Antonio , Pacciani Roberto , Scotti Del Greco Alberto , Spano Ennio
Published: November 2008

Numerical optimization techniques are increasingly used in the aerodynamic design of turbomachine blades. In the present paper, an existing three-dimensional high-lift turbine cascade was redesigned by means of CFD analyses and optimization techniques, based on the blade geometrical parameterization. A new parametric design tool was developed for this purpose. Blade geometry was handled in a fully three dimensional way, using Be´zier curves and surfaces for both camber-surface and thickness distribution. In the optimization procedure different techniques were adopted: a Genetic Algorithm (GA) strategy made it possible to considerably reduce two-dimensional profile losses, while the optimal stacking line was found based on a successive Design of Experiments (DOE) analysis. As a result, a new high-lift blade with higher performance was obtained; in addition, the effect of hub/tip leaning was presented and discussed.
Large-Scale Tests of Pyrotechnically Generated Aerosol Fire Extinguishing Systems for the Protection of Machinery Spaces and Gas Turbine Enclosures in Royal Naval Warships
Authors: Connell Martin , Glockling James L. D. , Neads Fiona
Published: November 2008

The first Royal Navy Destroyer for over 20 years is due to enter service in 2007. The base design for fire protection of the main machinery space (first shot) and Gas Turbine enclosure is by a carbon dioxide total flooding system. Ministry of Defence policy precludes the use Halon as part of the implementation of the Montreal Protocol on new construction ships. Carbon dioxide although the preferred solution for the T45 Destroyer is inherently toxic and places large demands on storage areas and makes this far from the ideal solution. Since 2001, the T45 Destroyer Warship Integration Team has been advancing the testing of two pyrotechnically generated aerosol systems as a replacement for carbon dioxide systems. Large-scale tests of aerosol agents were carried out at Darchem Flares IMO facility (internal dimensions 10m × 10m × 5m high) near Darlington, UK. The test programme was loosely based around the IMO (International Maritime Organisation) MSC/Circ. 1007 protocol but with greater amounts of clutter in the facility and more representative of a warship’s machinery space. Fire scenarios were also modified to make them more representative of naval circumstances with the pass criteria as specified in IMO MSC/Circ 1007 specification: • Class B fires extinguished within 30 seconds; • 15 minute hold period (no re-ignition); • fuel spray shut off 15 seconds after fire extinguishment; • at the end of hold period fuel spray should be restarted for 15 seconds with no re-ignition evident; • at the end of the test fuel trays must have sufficient fuel remaining to cover bottom of tray. In addition to the large-scale testing aerosol fire testing was conducted on a faithful replica of a Gas Turbine enclosure. The GT mock-up enclosure was designed to replicate in service thermal and airflow conditions within the gas turbine enclosure prior to the fire scenario and release of the agent. A hazard assessment was carried out by a focus group to determine expected fire conditions within the enclosure and these were replicated in a number of fire tests to evaluate the performance of the aerosols. Although derived from a common background and described by common terminology, very stark performance differences were found between the two systems tested. This paper will describe the test facility, tests carried out, performance result’s and proposed future use of pyrotechnic fire suppression aerosols within the Royal Navy.
Gas Turbine Industrial Fellowship Program
Authors: Day William H.
Published: November 2008

Under the Gas Turbine Industrial Fellowship Program, students in Bachelor’s, Master’s and Ph. D. programs studying gas turbine-related technology spend 10 to 12 weeks employed at the facilities of turbine manufacturers or users of gas turbine equipment. The program is funded by the U.S. Department of Energy. This paper describes the Fellowship program, its relationship to the DOE Turbine Program, the University Turbine Systems Research (UTSR) program, and plans for future Fellowship development.
Wear Resistant Polymer Matrix Composites for Aerospace Applications
Authors: Naik Subhash K. , Sutter James K. , Tabakoff Widen , Siefker Robert G. , Haller Harold S. , Cupp Randall J. , Miyoshi Kazuhisa
Published: November 2008

Polymer matrix composites (PMCs) are attractive for use in propulsion engine components due to their high specific strength. The use of composites could be even more advantageous if the sand erosion life of the component were extended, thereby reducing maintenance costs. NASA Glenn Research Center (NASA GRC) and Rolls-Royce Corporation have developed erosion resistant coatings that can extend PMC component life and are applicable to current available and advanced high temperature PMCs. This paper describes the performance of SANPRES and SANRES, two similar erosion resistant coating systems that were subjected to engine test conditions on Rolls-Royce AE 3007 engine bypass vanes.
Effects of Inlet Ramp Surfaces on the Aerodynamic Behaviour of Bleed Hole and Bleed Slot Off-Take Configurations
Authors: Leishman B. A. , Cumpsty N. A. , Denton J. D.
Published: November 2008

Bleed off-take air pressure and the interaction of the off-take with the primary flow through the blade passage is determined by: (1) the location of the bleed off-take at the endwall relative to the blade passage; (2) the bleed flow rate; and (3) the off-take geometry. In the companion paper (Leishman et al., 2004) the effect of bleed rate and endwall location was investigated using a circular hole bleed off-take configuration; the circular hole was tested at three endwall locations and for bleed flow rates between zero and 9 percent of the primary (core) flow through the blade passage. The effects of bleed off-take geometry are presented in this paper by comparing the aerodynamic behaviour of a number of generic bleed off-take configurations. Using results from low-speed cascade experiments and three-dimensional numerical calculations, the off-take configurations are compared with respect to the requirement to maximise bleed off-take air pressure and minimise loss generated within the blade passage. The off-take geometry, and especially the introduction of contoured inlet ramp surfaces to guide flow into the off-take for high bleed pressure, has a strong effect on its aerodynamic behaviour because it determines the extent to which flow within the off-take is coupled to the primary flow through the blade passage. In this paper, the off-take configurations that give the highest bleed pressure generally cause the highest levels of loss in the blade passage.
Predictions for the Effects of Freestream Turbulence on Turbine Blade Heat Transfer
Authors: Boyle R. J. , Ames Forrest E. , Giel P. W.
Published: November 2008

An approach to predicting the effects of freestream turbulence on turbine vane and blade heat transfer is described. Four models for predicting the effects of freestream turbulence were incorporated into a Navier-Stokes CFD analysis. Predictions were compared with experimental data in order to identify an appropriate model for use across a wide range of flow conditions. The analyses were compared with data from five vane geometries and from four rotor geometries. Each of these nine geometries had data for different Reynolds numbers. Comparisons were made for twenty four cases. Steady state calculations were done because all experimental data were obtained in steady state tests. High turbulence levels often result in suction surface transition upstream of the throat, while at low to moderate Reynolds numbers the pressure surface remains laminar. A two-dimensional analysis was used because the flow is predominantly two-dimensional in the regions where freestream turbulence significantly augments surface heat transfer. Because the evaluation of models for predicting turbulence effects can be affected by other factors, the paper discusses modeling for transition, relaminarization, and near wall damping. Quantitative comparisons are given between the predictions and data.
TMF Crack Initiation Lifing of Austenitic Carbide Precipitating Alloys
Authors: Hasselqvist Magnus
Published: November 2008

Recent work within DDIT has shown that Ni base superalloys like HAYNES230, Co base superalloys like HAYNES188, super stainless steels like HAYNES HR-120, and stainless steels like 253MA are similar from a materials modelling point of view. They are austenitic, delivered solutioned, and precipitate secondary carbides and other brittle phases in service and during cyclic tests at elevated temperature. These new phases result in a significantly reduced RT ductility, while the high temperature ductility is at most moderately reduced. Therefore, TMF cycles, which repeatedly go down to low temperatures, see an embrittled alloy whereas LCF tests at Tmax (in the TMF cycle) do not. This suggests that the classical use of LCF data at Tmax might given non-conservative life estimates. Literature studies and materials testing have confirmed that TMF data may be well below LCF data at Tmax verifying the non-conservatism of the classical methodology. Furthermore, the cyclic life tends to decrease with decreasing Tmin in TMF tests, and IP TMF is usually more detrimental than OP TMF due to creep-fatigue interaction. While standard TMF tests are closer to reality than LCF tests, we are still not certain that they capture all detrimental effects under component cycling, and are running additional, carefully planned, TMF tests on aged specimen at low Tmin values to improve the analysis. More tests, especially biaxial IP TMF tests, will eventually be needed to get a comprehensive picture. A new TMF data backed model has, however, already shown a higher precision when compared with service experience than the classical creep-fatigue methodology which is calibrated with LCF data at Tmax. Further testing and analysis will enable us to refine the TMF model and extend it to additional ACP alloys. The main input to the TMF model is the stabilised inelastic strain range, as calculated by the constitutive model described in an earlier ASME Turbo paper, GT2002-30659.
Auditing Supports the Integration of Management Systems in the Nuclear Industry
Authors: Beckmerhagen I. A. , Berg H. P. , Karapetrovic S. V. , Willborn W. O.
Published: November 2008

Integration of function-specific management systems in organizations is rapidly becoming a topic of interest for managers and auditors alike. This is mainly due to the proliferation of management system standards that foster compliance with the stated criteria for quality, environmental, occupational health and safety, social responsibility and other function-specific aspects of performance. While most of the available literature on this topic focuses on the integration of standards, there is comparatively little information available on how to actually build an integrated system internally. This paper hypothesizes that, besides using audits for the implementation of the available procedures, audits can provide an excellent basis for these integration efforts. Therefore the prerequisites, strategies and resources necessary for an effective audit in support of integrated management systems are discussed. The paper also describes how audits are used to improve a combined quality and safety management system in a German nuclear facility.
Nuclear Power: Time To Start Again
Authors: Rezak William D.
Published: November 2008

One of America’s best kept secrets is the success of its nuclear electric power industry. This paper presents data which support the construction and operating successes enjoyed by energy companies that operate nuclear power plants in the US. The result—the US nuclear industry is alive and well. Perhaps it’s time to start anew the building of nuclear power plants. Let’s take the wraps off the major successes achieved in the nuclear power industry. Over 20% of the electricity generated in the United States comes from nuclear power plants. An adequate, reliable supply of reasonably priced electric energy is not a consequence of an expanding economy and gross national product; it is an absolute necessity before such expansion can occur. It is hard to imagine any aspect of our business or personal lives not, in some way, dependent upon electricity. All over the world (in over 30 countries) nuclear power is a low-cost, secure, safe, dependable, and environmentally friendly form of electric power generation. Nuclear plants in these countries are built in six to eight years using technology developed in the US, with good performance and safety records. This treatise addresses the success experienced by the US nuclear industry over the last 40 years, and makes the case that this reliable, cost-competitive source of electric power can help support the economic engine of the country and help prevent experiences like the recent crises in California and the Northeast. Traditionally, the evaluation of electric power generation facility performance has focused on the ability of plants to produce at design capacity for high percentages of the time. Successful operation of nuclear facilities is determined by examining capacity or load factors. Load factor is the percentage of design generating capacity that a power plant actually produces over the course of a year’s operation. This paper makes the case that these operating performance indicators warrant renewed consideration of the nuclear option. Usage of electricity in the US now approaches total generating capacity. The Nuclear Regulatory Commission has pre-approved construction and operating licenses for several nuclear plant designs. State public service commissions are beginning to understand that dramatic reform is required. The economy is recovering and inflation is minimal. It’s time, once more, to turn to the safe, reliable, environmentally friendly nuclear power alternative.
Case Study on the Role of Socio-Technical Influences on the Implementation and Success of Nuclear Power in France
Authors: Jedani Tony
Published: November 2008

To fully understand a technological development one must appreciate social, political and economic factors in addition to the technological components (Hughes, 1991). The systems perspective, asserted by Hughes, implies that technologies cannot be understood in isolation, but only in their contexts, especially in their systemic contexts. This theory is illustrated through an examination of France’s implementation of its nuclear power program in the early 1970’s. Nuclear power provided France with the opportunity to achieve energy independence and, as a result, political control over its energy supply. The scope of this case study is limited to consideration of the socio-technical influences on the rise of nuclear power in France and includes an examination of the technical aspects of the innovation. In considering the socio-technical system encompassing France’s adoption of nuclear power, this case study will contemplate: how France was able to persuade its people to accept nuclear power; what it is about French culture and politics that allowed them to succeed where most other countries have failed; the break throughs that led to the broad commercialisation of nuclear power in France in the 1970’s; and how France achieved its status as one of the world’s top producers of nuclear energy. The time period during which this study is based is limited to the early 1970’s, when France was reliant upon external energy supplies, up until the present day, where nuclear power has become France’s main source of energy, thus contributing to France’s autonomy in terms of its energy supply. This study will not address the issue of nuclear waste or the nuclear power safety debate which is also beyond the scope of this study.
Risk-Informed Preventive Maintenance Optimization
Authors: Liming James K. , Salter James E.
Published: November 2008

Past and ongoing electric generating station owner investments in plant information technology (such as database query applications and other client workstation tools) have made it possible for plant staffs to utilize information contained in the work management systems to quickly link equipment failure modes to related preventative maintenance (PM) activities. A typical pressurized water reactor feedwater (FW) system is applied as the “target system” for examples in this paper. This typical FW system is comprised of approximately 3,800 “tag” or “part number” items which in turn represent about 16,300 failure modes. Effective risk-informed asset management (RIAM) of FW preventive maintenance (PM) activities requires these failure modes to be modeled in a plant availability model. In this paper we present development of a process for supporting PM optimization, applying cost-benefit-risk analysis and RIAM tools and techniques. In this preventive maintenance optimization (PMO) process, PM activities are evaluated for their projected impacts on plant profitability and nuclear safety. PM activities (PMs) are “optimized” for desirable impact to help ensure electric utilities maintain or improve upon high levels of nuclear safety and profitability. In this PMO application the level of detail of the target system(s) is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture. Results of case studies in FW system PMO using typical plant data are presented.
Probabilistic Safety Assessment Technology for Commercial Nuclear Power Plant Security Evaluation
Authors: Liming James K. , Johnson David H. , Dykes Andrew A.
Published: November 2008

Commercial nuclear power plant physical security has received much more intensive treatment and regulatory attention since September 11, 2001. In light of advancements made by the nuclear power industry in the field of probabilistic safety assessment (PSA) for its power plants over that last 30 years, and given the many examples of successful applications of risk-informed regulation at U. S. nuclear power plants during recent years, it may well be advisable to apply a “risk-informed” approach to security management at nuclear power plants from now into the future. In fact, plant PSAs developed in response to NRC Generic Letter 88-20 and related requirements are used to help define target sets of critical plant safety equipment in our current security exercises for the industry. With reasonable refinements, plant PSAs can be used to identify, analyze, and evaluate reasonable and prudent approaches to address security issues and associated defensive strategies at nuclear power plants. PSA is the ultimate scenario-based approach to risk assessment, and thus provides a most powerful tool in identifying and evaluating potential risk management decisions. This paper provides a summary of observations of factors that are influencing or could influence cost-effective or “cost-reasonable” security management decision-making in the current political environment, and provides recommendations for the application of PSA tools and techniques to the nuclear power plant operational safety response exercise (OSRE) process. The paper presents a proposed framework for nuclear power plant probabilistic terrorist risk assessment (PTRA) that applies these tools and techniques.
Analysis of Magnetohydrodynamic Flow in Annular Duct
Authors: Yoo Geun Jong , Choi Hoon Ki , Eun Jae Jeong
Published: November 2008

Numerical analysis is performed for magnetic and magnetohydrodynamic (MHD) flow fields in electromagnetic (EM) pump. A finite volume method is applied to solve governing equations of magnetic field and the Navier-Stokes equations. Vector and scalar potential methods are adopted to obtain the electric and magnetic fields and the resulting Lorentz force in solving Maxwell equations. The magnetic field and velocity distributions are found to be affected by the phase of applied electric current and the magnitude of the Reynolds number. Computational results indicate that the magnetic flux distribution with changing phase of input electric current is characterized by pairs of counter-rotating closed loops. The axial velocity distributions are represented with S-type profiles for the case of the r-direction of Lorentz force dominated flows.
Investigation of a Possible Emergency Procedure for the VVER 1000 NPP in Case of a Total Loss of Feedwater and a Main Steam Line Break
Authors: Muellner Nikolaus , Giannotti Walter , D’Auria Francesco
Published: November 2008

Accident management procedures associated with nuclear power plant beyond design basis accidents should be developed with the aid of the more recent versions of advanced computational tools (best estimate codes) in order to verify the effectiveness of normal operation systems of the plant to avoid or minimize core damage. A. Madeira showed in her paper “A PWR Recovery Option for a Total Loss of Feedwater Beyond Design Basis Scenario” that it is possible to safely control a total loss of feedwater scenario in the Angra2 nuclear power plant, using two emergency procedures, namely the opening of the steam generator (SG) relief valves, and on the primary side the complete manual opening of all pressurizer relief and safety valves. This paper investigates the effectiveness of the procedure opening of the SG relief valves, followed by primary side feed and bleed for a generic VVER-1000 NPP in case of a total loss of feed water. The results indicate that the procedure is successful in reducing the primary side pressure and temperature to safe conditions, i.e. long term core cooling is achievable.
A Reliability Study of the Diesel Generator of the HANARO Research Reactor
Authors: Kim Hyungkyoo , Jung Hoansung , Wu Jongsup , Jun Byungjin
Published: November 2008

The diesel generator of HANARO provides Class III (standby) power during the loss of Class IV (off site) power events, for the important reactor system equipment such as the emergency water supply system, emergency ventilation system, etc. The HANARO has been stabilizing its operation since 1995, so the diesel generator failure data from 1995 to the end of 2002 has been collected. This study shows the failure-to-start probabilities and the failure-to-run probabilities of diesel generator within the 90% confidence bounds. A comparison with the NUREG and AECL studies is shown in this paper.
Application of Power Reactors for Transmutation of Actinides
Authors: Bergelson B. R. , Gerasimov A. S. , Tikhomirov G. V.
Published: November 2008

Heavy-water and light-water power reactors can be used for partial transmutation of radwaste. Such transmutation allows us to limit on relatively low level the radiotoxicity accumulated in long-term storage of spent fuel. At transmutation in PHWR-880 reactor operating in the mode of self-service, equilibrium radiotoxicity in storage facility and the time of its achievement are 4–5 times less than for the transmutation in power reactor VVER-1000.
Bifurcation Analysis Using the System Code RAMONA
Authors: Dokhane A. , Chawla R. , Hennig D. , Rizwan-Uddin
Published: November 2008

This paper presents a study in which the bifurcation analysis is carried out using the system code RAMONA. As a first step, a correspondence hypothesis is proposed to underline the relationship between a stable (unstable) limit cycle solution and the occurrence of supercritical (subcritical) Hopf bifurcation in the modeling of boiling water reactor stability behaviour. A detailed local bifurcation analysis is then carried out in the narrow environment of one representative operational point for the Leibstadt NPP. Based on the experience acquired during stability and bifurcation analyses using our reduced order model, the results found using RAMONA could be explained. As a result, this is the first time that the occurrence of a subcritical Hopf bifurcation has been identified using RAMONA, or any other system code.
25-Years Three-Fluid Modeling Experience: Successes and Limits
Authors: Kolev Nikolay Ivanov
Published: November 2008

The paper presents many examples of successful use of the three fluid processes modeling in 1D-networks, 3D-single volumes and 3D-boundary fitted volumes in the nuclear engineering. The modeling techniques, which have proof to be useful, are underlined and recommended for future use. The limits of the method are elaborated and recommended for future investigation.
External Cooling: The SWR 1000 Severe Accident Management Strategy
Authors: Kolev Nikolay Ivanov
Published: November 2008

This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have been determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool.
The Use of a Polyurea Coating as a Replacement for IP-2 Containers in the Transportation of Large, Low-Level Waste Material
Authors: Laffitte John , Lagos Leo
Published: November 2008

The decontamination and decommissioning of the Department of Energy Rocky Flats site has resulted in a number of large waste items that need to be shipped to waste collection areas. However, in order to transport these items, they need to be size-reduced and packed into DOT-approved containers such as standard industrial packaging (IP) waste containers. The size reduction process costs time and money while exposing workers to potential health and safety risks. Spray-on polyurea coatings have the potential of allowing large items to be encapsulated by the coating and transported without having to size-reduce the items. The Hemispheric Center for Environmental Technology (HCET) at Florida International University (FIU) recently performed an engineering analysis that includes mechanical testing and small-scale drop tests to determine the feasibility of using a spray-on polyurea coating for shipment of large, low-level waste items instead of using standard IP waste containers.
Study of Thermal Influence on Tubes Due to Sodium-Water Reactions in LMFBR Steam Generator
Authors: Hamada Hirotsugu , Kurihara Akikazu , Nishimura Masahiro
Published: November 2008

A study of thermal influence on heat-transfer tubes in sodium-water reactions is carried out to evaluate the tube rupture due to overheating in the water leak accident of an LMFBR steam generator (SG). By assuming the sodium-water reaction jet to be a two-phase flow that consists of sodium and hydrogen, the heat-transfer characteristics are examined and a simple model of effective heat-transfer coefficient (HTC) is proposed for the safety evaluation of the SG. Comparison of the model with experimental data leads to the following conclusions: An upper limit exists in the HTC between reaction jet and tube wall, and it is equivalent in approximation to the HTC of single-phase sodium flow. The HTC can be written in simple form as functions of the HTC of single-phase sodium flow, void fraction and temperatures of sodium, hydrogen and tube wall. Hydrogen provides negligible heating effect, so that the apparent HTC would decrease with increase of the hydrogen temperature that can readily surpass that of sodium. The outer-surface temperature of tube wall would not rise so high beyond the temperature of sodium that is excellent in heat-transfer characteristics, even if tube wall is exposed to the high-temperature hydrogen. The transient heat conduction analysis with the mean value of the data can appropriately evaluate the outer-surface temperature of tube wall by the metallographic observation, while the analysis with the maximum value can conservatively evaluate the tube wall temperature.
System Design Study of a Membrane Reforming Hydrogen Production Plant Using a Small Sized Sodium Cooled Reactor
Authors: Chikazawa Yoshitaka , Hori Toru , Konomura Mamoru , Sato Hiroyuki , Uchida Shouji
Published: November 2008

In phase II of a feasibility study on commercialized fast reactor cycle system of Japan Nuclear Cycle Development Institute, we are finding a concept of a multi-purpose small sized fast reactor with various requirements, such as, safety and improved economical competitiveness. In this study, a membrane reforming hydrogen production plant using a small sized sodium cooled reactor was designed as one of promising concepts. In the membrane reformer, methane and steam are reformed into carbon dioxide and hydrogen with sodium heat at a temperature 500deg-C. In the equilibrium condition, steam reforming proceeds with catalyst at a temperature more than 800deg-C. Using membrane reformers, the steam reforming temperature can be decreased from 800 to 500deg-C because the hydrogen separation membrane remove hydrogen selectively from catalyst area and the partial pressure of hydrogen is kept much lower than equilibrium condition. In this study, a hydrogen and electric co-production plant has been designed. The reactor thermal output is 375MW and 25% of the thermal output is used for hydrogen production (70000Nm3/h). The hydrogen production cost is estimated but it is still higher than the economical goal. The major reason of the high cost comes from the large size of hydrogen separation reformers because of the limit of hydrogen separation efficiency of palladium membrane. A new highly efficient hydrogen separation membrane is needed to reduce the cost of hydrogen production using membrane reformers. There is possibility of multi-tube failure in the membrane reformers. In future study, a design of measures against tube failure and elemental experiments of reaction between sodium and reforming gas will be needed.
Application of the PDM (Point Defect Model) to the Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors
Authors: Park Pilyeon , Urquidi-Macdonald Mirna , Macdonald Digby D.
Published: November 2008

The PDM [Point Defect Model, D. D. Macdonald, Pure Appl. Chem., 71, 951 (1999)] describes the corrosion of passive metals in aqueous media in terms of the generation and annihilation of point defects at the passive film interfaces. In the current work, we have modified the PDM to provide a comprehensive, atomic scale description of the growth of bilayer passive films on zirconium to simulate the corrosion of Zircaloy fuel cladding in BWRs and PWRs under high burn-up conditions. Two models have been formulated; one comprising a hydride inner (barrier) layer and an oxide outer layer and other comprising an oxide inner layer and an oxide outer layer for PWR and BWR cladding, respectively. Since there are currently no experimental data for the kinetics of defect generation and annihilation at the passive film interfaces for Zircaloys under PWR/BWR conditions, of the type that are required for this analysis, this paper focuses only on exploring and predicting trends in the corrosion behavior of Zircaloy by using prototypical values for various electrochemical parameters. We derive equations for predicting the barrier layer thickness as a function of the applied voltage, pH, porosity, and temperature for both BWR and PWR primary water chemistry conditions.
Reliability Analysis of Digital Reactor Protection System
Authors: Yamashita Masahiro , Miura Satoshi , Fukuda Mamoru , Hirano Mitsumasa
Published: November 2008

The reliability analysis of the digital reactor protection system (RPS) is one of the essential parts in the probabilistic safety assessment (PSA) of the advanced boiling water reactor (ABWR). In this study, the reliability model and methodology were modified to evaluate the reliability of the digital RPS installed in the Japanese ABWR plant. The hardware failure rates in the foreign data source of digital components were applied, based on the similarity of the function of the digital components. The hardware failure rates of the digital components were estimated to range from 1.0E−5 (/hr) to 1.0E−7 (/hr), according to the types of the components. The software error events and their recovery factors in the design and fabrication stages were evaluated, considering the verification and validation process provided by the Japanese industry guideline on the digital reactor protection system. Then, the software failure probability of the programmable digital component was evaluated, utilizing the probability of software error events and their recovery factors. The software failure probability was estimated to be 3.3E−7 (/demand), which was about one order higher than that of our previous estimation. These models and results were applied to evaluate the reactor trip system (RTS) and the engineered safety feature (ESF) actuation system of the ABWR plant, both of which are the subsystems of the RPS. The unavailability of the digital RTS was estimated to be the mean value of 7.2E−06 (/demand). If both an alternate rod insertion (ARI) and a manual scram were considered, the unavailability was estimated to decrease to 1.6E−09. This value was nearly equal to the mean value of the previous study, 1.1E−09 (/demand), even though the quantification model and data were considerably modified, including the software failure probability. The system unavailability of the emergency core cooling system (ECCS) was also evaluated in conjunction with the ESF actuation system, in order to investigate the effect of the model and data modification. The ECCS unavailability was estimated to be also nearly equal to the same values as the previous estimation, because the system unavailability was dominated by the unavailability of the mechanical components, such as pumps, valves, etc. The sensitivity analyses were conducted systematically, in order to evaluate the effect of the modeling uncertainty on the digital RTS unavailability. The results indicated that the unavailability of the digital RTS only changed within the range of factor 2, even though the various assumptions were used on the hardware and the software failure of the digital components.
Preliminary Applications of the New NEPTUNE Two-Phase CFD Solver to Pressurized Thermal Shock Investigations
Authors: Boucker Marc , Lavie´ville Je´roˆme , Martin Alain , Be´chaud Ce´line , Bestion Dominique , Coste Pierre
Published: November 2008

The objective of this communication is to present some preliminary applications to pressurized thermal shock (PTS) investigations of the CFD two-phase flow solver of the new NEPTUNE thermal-hydraulics platform, being jointly developed by EDF (Electricite´ De France) and CEA (Commissariat a` l’Energie Atomique) and also supported by IRSN (Institut de Radioprotection et Suˆrete´ Nucle´aire) and FRAMATOME-ANP. In the framework of plant life extension, the Reactor Pressure Vessel (RPV) integrity is a major concern, and an important part of RPV integrity assessment is related to PTS analysis. In the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur, such as condensation due to the Emergency Core Cooling (ECC) injections of sub-cooled water. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. In that purpose, a program has been set up to extend the capabilities of the NEPTUNE two-phase CFD solver. The major challenge is to develop new physical models to take into account the complex two-phase phenomena: free surface, jet of cold water through steam, entrainment of droplets by steam, entrainment of bubbles by water, and, above all direct contact condensation. Turbulent mixing in the liquid layer plays a central role by controlling at the same time the thermal mixing in the water and the condensation rate at the interface between liquid and vapor. A sustained effort in this area has started both at CEA and EDF. New models are developed and validated with the help of available experimental data sets, which are unfortunately very few. Beyond analytical single-effect cases, it is important to compute instrumented cases close to the life-size one, including major physical effects. This is the case of the COSI experimental facility, representing a cold leg with ECC injection, where temperature profiles have been measured. Numerical results on a steady-state COSI test are presented and analyzed, showing encouraging results. However, research work is still required to achieve a better modeling of the heat transfer and turbulence at the steam-water interface.
Projection of the Neutronic and Thermal Fuel Rod Behavior in a BWR
Authors: Hernandez Lopez Hector , Ortiz Villafuerte Javier
Published: November 2008

Currently, at the Instituto Nacional de Investigaciones Nucleares (National Institute for Nuclear Research) in Mexico, it is being developed a computational code for evaluating the neutronic, thermal and mechanical performance of a fuel element at several different operation conditions. The code is referred as to MCTP (Multigrupos con Temperaturas y Potencia), and is benchmarked against data from the Laguna Verde Nuclear Power Plant (LVNPP). In the code, the neutron flux is approximated by six groups of energy: one group in the thermal region (E < 0.625 eV), four in the resonances region (0.625 eV < E < 0.861 MeV), and one group in the fast region (E > 0.861 MeV). Thus, the code is able to determine the damage to the cladding due to fast neutrons. The temperature distribution is approximated in both axial and radial directions taking into account the changes in the coolant density, for both the single and two-phase regions in a BWR channel. It also considerate the changes in the thermal conductivity of all materials involved for the temperature calculations, as well as the temperature and density effects in the neutron cross sections. In the code, fuel rod burnup is evaluated. Also, plutonium production and poison production from fission. In this work, the neutronic and thermal performance of fuel rods in a 10×10 fuel assembly is evaluated. The fuel elements have a content of 235U. The fuel assembly was introduced to the unit 1 of LVNPP reactor core in the cycle 9 of operation, and will stay in during three cycles. In the analysis of fuel rod performance, the operating conditions are those for the cycle 9 and 10, whereas for the current cycle (cycle 11) the reactor is projected to operate during 460 days. The analysis for cycle 11 uses the actual location of the fuel assembly that will have in the core. The results show that the fuel rods analyzed did not reach the thermal limits during the cycles 9 and 10, as expected, and for cycle 11 the same thermal limits are not predicted to be reached.
CAREM Concept: A Competitive SMR
Authors: Delmastro Dari´o , Gime´nez Marcelo , Florido Pablo C. , Daverio Hernando , Serra Oscar , Blanco Ani´bal , Mueller Pablo
Published: November 2008

CAREM concept is based in a cost effective primary system configuration. The use of an integrated primary cooling system, self-pressurization, internal control rod drives and boron free coolant are many of the distinctive features that characterize the system. In this paper CAREM is presented. This is an advanced reactor concept with new engineering solutions based in specific developments, the mature engineering of the prototype CAREM-25 and the wide worldwide experience in safe LWR operation.