bluuebunny/crossref_metadata_2025 huggingface dataset:
Ongoing Development and Recent Results of the Research in the Swedish Gas Turbine Centre (GTC)
Authors: Sundkvist Sven Gunnar , Andersson Michael , Gherman Bogdan , Sveningsson Andreas , Vogt Damian
Published: November 2008

This paper describes a way of co-operation between industries, universities and government that has proven to be very fruitful. The Swedish Gas Turbine Centre (GTC) is constituted as a research consortium between technical universities and gas turbine industry. The overall goal of the centre, that was founded in 1996 on a governmental initiative, is to build up a basis of knowledge at Swedish universities to support the industrial development in Sweden of gas turbines of the future with expected requirements on low emissions, high efficiencies, high availability, and low costs. Since the start the research has had a focus on high temperature components of gas turbines (combustion chamber and turbine). This is also reflected in the on-going development phase where the research program consists of four project areas: cooling technology, combustion technology, aeroelasticity, and life time prediction of hot components. The projects are aiming at developing design tools and calculation and verification methods within these fields. A total of eleven research students (among them one industrial PhD student) are active in the centre at present. Numerical analysis as well as experimental verification in test rigs are included. The program has so far produced eleven Licentiate of Engineering and five PhD. On-going activities and recent results of the research in the four research areas are presented: • A new test rig for investigation of time-dependent pressures of three-dimensional features on a vibrating turbine blade at realistic Mach, Reynolds and Strouhal numbers and first experimental results. • Results of numerical simulations of heat loads on turbine blades and vanes, especially platform cooling. • First results of numerical investigations of combustion and thermo-acoustic instabilities in gas turbine chambers. • Experimental investigation of crack propagation in gas turbine materials using the scanning electron microscope (SEM).
Effects of Bleed Rate and Endwall Location on the Aerodynamic Behaviour of a Circular Hole Bleed Off-Take
Authors: Leishman B. A. , Cumpsty N. A. , Denton J. D.
Published: November 2008

In a jet engine bleed off-takes on the hub and casing endwalls, part-way through the compressor, supply high-pressure air for cooling, sealing, de-icing and aircraft cabin air applications; bleed also assists compressor operation at part-speed conditions. Two separate issues are of interest: the bleed off-take air pressure and the interaction of the bleed off-take with the primary flow through the blade passage. In this paper, the aerodynamic behaviour is presented for a circular-hole bleed off-take at three endwall locations within a stationary cascade blade passage: at mid-passage; near the blade pressure-surface; and near the blade suction-surface. Results from low-speed cascade experiments are complemented by three-dimensional numerical calculations using an unstructured mesh-based solver, in which the blade passage and bleed off-take geometry are fully modelled. The bleed off-take location and the magnitude of bleed rate influence the off-take air pressure and the interaction with the primary passage flow. For optimum design at zero and low bleed rates, off-takes near the blade pressure-surface give the highest bleed air pressures and minimum loss in the blade passage. For minimum blade passage loss at higher bleed rates, however, it is necessary to take bleed near the blade suction-surface. The paper discusses the causes for this pattern of behaviour.
GTPOM: Thermo-Economic Optimization of Whole Gas Turbine Plant
Authors: Knight Richard , Obana Mitsuru , von Wowern Christer , Mitakakis Athanasios , Perz Erhard , Assadi Mohsen , Mo¨ller Bjo¨rn F. , Sen Pratyush , Potts Ian , Traverso Alberto , Torbidoni Leonardo
Published: November 2008

Trends towards distributed power generation and the deregulation of energy markets are increasing the requirement for software tools that optimize power generation plant design and operation. In this context, this paper describes the GTPOM (thermo-economic optimization of whole gas turbine plant) European project, funded in part through the European Commission’s 5th Framework Programme, focusing on the development and demonstration of an original software tool for the thermo-economic analysis and optimization of conventional and advanced energy systems based on gas turbine plant. PSEconomy, the software tool developed during the GTPOM project, provides a thermo-economic optimization capability for advanced and more-conventional energy systems, enabling the complex trade-offs between system performance and installed costs to be determined for different operational duties and market scenarios. Furthermore, the code is capable of determining the potential benefits of innovative cycles or layout modifications to existing plants compared with current plant configurations. The economic assessment is performed through a complete through-life cycle cost analysis, which includes the total capital cost of the plant, the cost of fuel, O&M costs and the expected revenues from the sale of power and heat. The optimization process, carried out with a GA-based algorithm, is able to pursue different objective functions as specified by the User. These include system efficiency, through-life cost of electricity and through-life internal rate of return. Three case studies demonstrating the capabilities of the new tool are presented in this paper, covering a conventional combined cycle system, a biomass plant and a CO2 sequestration gas turbine cycle. The software code is now commercially available and is expected to provide significant advantages in the near and long-term development of energy cycles.
Design of Ceramic Components for an Advanced Micro-Turbine Engine
Authors: Tredway Bill , Shi Jun , Holowczak John , Vedula Venkata , Bird Connie E. , Ochs S. Scott , Bertuccioli Luca , Bombara David J. , Green Kevin E.
Published: November 2008

Ceramic components, due to their high temperature capability, allow significantly higher turbine inlet temperatures with minimal cooling. Hot-section engine components, including combustor, integral vane ring, integrally bladed turbine rotor, and turbine tip shroud were designed for an advanced micro-turbine engine, with special attention to attachment methods that minimize thermal stresses due to large differences between coefficients of thermal expansion between metallic and ceramic materials. Detailed aerodynamic, thermal and stress analyses were performed. Both steady state and transient conditions were evaluated to guide design decisions that lead to optimal component reliability and manufacturability. This paper describes the component design, analysis, and fabrication experiences with silicon based monolithic ceramic materials.
Effects of Dimple Depth on Nusselt Numbers and Friction Factors for Internal Cooling in a Channel
Authors: Burgess N. K. , Ligrani P. M.
Published: November 2008

Experimental results, measured on dimpled test surfaces placed on one wall of different channels, are given for a ratio of air inlet stagnation temperature to surface temperature of approximately 0.94, and Reynolds numbers based on channel height from 9,940 to 74,800. The data presented include friction factors, local Nusselt numbers, spatially-averaged Nusselt numbers, and globally-averaged Nusselt numbers. The ratios of dimple depth to dimple print diameter δ/D are 0.1, 0.2, and 0.3 to provide information on the influences of dimple depth. The ratio of channel height to dimple print diameter is 1.00. At all Reynolds numbers considered, local and spatially-resolved Nusselt number augmentations increase as dimple depth increases (and all other experimental and geometric parameters are held approximately constant). These are attributed to: (i) increases in the strengths and intensity of vortices and associated secondary flows ejected from the dimples, as well as (ii) increases in the magnitudes of three-dimensional turbulence production and turbulence transport. The effects of these phenomena are especially apparent in local Nusselt number ratio distributions measured just inside of the dimples, and just downstream of the downstream edges of the dimples. Data are also presented to illustrate the effects of Reynolds number, and streamwise development for δ/D = 0.1 dimples. Significant local Nusselt number ratio variations are observed at different streamwise locations, whereas variations with Reynolds number are mostly apparent on flat surfaces just downstream of individual dimples.
A Transient Infrared Thermography Method for Simultaneous Film Cooling Effectiveness and Heat Transfer Coefficient Measurements From a Single Test
Authors: Ekkad Srinath V. , Ou Shichuan , Rivir Richard B.
Published: November 2008

In film cooling situations, there is a need to determine both local adiabatic wall temperature and heat transfer coefficient to fully assess the local heat flux into the surface. Typical film cooling situations are termed three temperature problems where the complex interaction between the jets and mainstream dictates the surface temperature. The coolant temperature is much cooler than the mainstream resulting in a mixed temperature in the film region downstream of injection. An infrared thermography technique using a transient surface temperature acquisition is described which determines both the heat transfer coefficient and film effectiveness (non-dimensional adiabatic wall temperature) from a single test. Hot mainstream and cooler air injected through discrete holes are imposed suddenly on an ambient temperature surface and the wall temperature response is captured using infrared thermography. The wall temperature and the known mainstream and coolant temperatures are used to determine the two unknowns (heat transfer coefficient and film effectiveness) at every point on the test surface. The advantage of this technique over existing techniques is the ability to obtain the information using a single transient test. Transient liquid crystal techniques have been one of the standard techniques for determining h and η for turbine film cooling for several years. Liquid crystal techniques do not account for non uniform initial model temperatures while the transient IR technique measures the entire initial model distribution. The transient liquid crystal technique is very sensitive to the angle of illumination and view while the IR technique is not. The IR technique is more robust in being able to take measurements over a wider temperature range which improves the accuracy of h and η. The IR requires less intensive calibration than liquid crystal techniques. Results are presented for film cooling downstream of a single hole on a turbine blade leading edge model.
CFD Modeling of the Atomization of Plain Liquid Jets in Cross Flow for Gas Turbine Applications
Authors: Khosla Sachin , Crocker D. Scott
Published: November 2008

A numerical model for liquid jet atomization in a subsonic gas cross flow has been developed and incorporated into a CFD code. The model is designed primarily for the shear breakup regime, which is appropriate for many fuel injection applications. The model considers Weber number and momentum flux ratio ranges that are dominated by either jet surface breakup or column breakup. A boundary layer stripping model has been modified to account for both shearing from the column and shear primary breakup of large drops. Further secondary breakup was modeled with the Rayleigh-Taylor model. The effect of drop distortion on the drag is also considered. Results of the model have been compared with experimental data for jet-A liquid jets in air cross flows with varying pressure, air velocity, and liquid-to-gas momentum flux ratio. Comparisons were made for drop volume flux and drop size as a function of distance from the injector wall. Trends were captured for liquid penetration associated with varying momentum flux ratio, and for drop size as a function distance from the wall. In general, agreement between measurements and CFD predictions were quite good. Areas of disagreement could be reasonably explained by the model’s inherent inability to capture the wake flow behind the liquid column.
A Computational Study of the Effect of Fuel Blends on Pollutants Emissions in Flames
Authors: Naha Sayangdev , Aggarwal Suresh
Published: November 2008

Requirement of significantly reducing NOx and particulate emissions while maintaining combustor performance is one of the main drivers for combustion research. Fuel blending represents a very promising approach for reducing both NOx and particulate emissions from flames. This paper reports an investigation on the effects of using hydrogen blending with different fuels on pollutants emission in counterflow flames. The fuels investigated include methane and n-heptane. Methane-hydrogen flame is computed using the GRI-3.0 mechanism, while n-heptane-hydrogen flame is computed by combining the Held’s oxidation mechanism with the Li and Williams’ NOx mechanism. Results indicate that in methane/hydrogen blends, emission of NOx is increases with hydrogen addition, but the effect is not significant. However, emission of CO and C2H2 (which is a very important soot precursor) are affected significantly by the addition of hydrogen. First, addition of hydrogen decreases the carbon content in the fuel for the same strain rate. Second, the chemistry of combustion also changes due to higher reactivity and higher concentration of H, O and OH radicals produced due to hydrogen addition. In heptane/hydrogen blends, all the three pollutant species investigated (NO, CO and C2H2) are found to be affected significantly by hydrogen addition. The effect of pressure on pollutants emission has also been investigated.
Evaluation of Operating Experience for Early Recognition of Deteriorating Safety Performance
Authors: Beckmerhagen I. A. , Berg H. P.
Published: November 2008

One of the most difficult challenges facing nuclear power plants is to recognize the early signs of degrading safety performance before regulatory requirements are imposed or serious incidents or accidents occur. Industry experience (also outside nuclear industry) has shown that these are typical issues which should be considered when looking for such early warning signs. Therefore, it is important that nuclear power plant operators have the capability to trend, analyse and recognize early warning signs of deteriorating performance. It is necessary that plant operators are sensitive to these warning signs which may not be immediately evident. Reviewing operating experience is one of the main tasks for plant operators in their daily activities. Therefore, self assessment should be at the centre of any operational safety performance programme. One way of applying a self assessment program is through the following four basic elements: operational data, events, safety basis, and related experience. This approach will be described in the paper in more detail.
Assessment of MELCOR 1.8.5 Versus Different Versions of SCDAP/RELAP5 MOD 3.3 With Lower Head Creep Rupture Analysis of Alternative Accident Sequences of the Three Mile Island Unit 2
Authors: Chan Sang Lung
Published: November 2008

The objective of this analysis is to assess MELCOR 1.8.5-RG against SCDAP/RELAP5 MOD 3.3kz (SR5m33kz), and SCDAP/RELAP5 MOD 3.3bf (SR5m33bf). This lower head creep rupture analysis considers: (1) Three Mile Island Unit 2 (TMI-2) alternative accident sequence-1, and (2) TMI-2 alternative accident sequence-2. SCDAP/RELAP5 model of TMI-2 alternative accident sequence-1 includes the continuation of the base case of the TMI-2 accident with the reactor coolant pumps (RCP) tripped, and the High Pressure Injection System (HPIS) throttled after approximately 6000 s accident time, SCDAP/RELAP5 model of TMI-2 alternative accident sequence-2 is derived from the TMI-2 base case accident by tripping the RCP after 6000 s, and the HPIS is reactivated after 12,012 s. MELCOR model of TMI-2 alternative accident sequence-1 is based on MELCOR TMI-2 phase-2 model by tripping the RCP and throttling back the makeup flows to zero from 6000 s onward. In MELCOR model of TMI-2 alternative accident sequence-2, the RCP are tripped from 6000 s and the constant makeup flow rate of 3.75 kg/s — including pump seal flow rate, but without HPIS flow rate — is activated from 6000 s and beyond 10440 s. The simulation is run until the lower head wall ruptures. In addition, the lower head penetration failure is also calculated with MELCOR for both TMI-2 alternative accident sequences. Lower head temperature contours calculated with SCDAP/RELAP5 are visualized and animated with open source visualization freeware ‘OpenDX’. Significant findings of the analysis include: (1) the TMI-2 lower head wall fails by creep rupture with either deactivations or activations of the HPIS; (2) for the TMI-2 alternative accident sequence-1 the time to creep rupture calculated with MELCOR 1.8.5-RG, SR5m33kz, and SR5m33bf agrees reasonably; (3) the calculation with MELCOR for the TMI-2 alternative accident sequence-1 predicts that the lower head wall failure occurred earlier than penetration failure, while MELCOR predicts the opposite for the TMI-2 alternative accident sequence-2; (4) calculation with MELCOR for TMI-2 alternative accident sequence-2 shows that when the lower head wall fails the temperature calculated with MELCOR is 1810.9 K, which exceeds the melting temperature of 1789 K for carbon steel; (5) calculations with both SR5m33kz and SR5m33bf for both TMI-2 alternative accident sequences indicate that different lower head wall locations fail rapidly one after another by a delay of a few seconds, while this is not the case for MELCOR.
One Dimensional Thermal-Hydraulic Stability Analysis of Supercritical Fluid Cooled Reactors
Authors: Zhao Jiyun , Saha Pradip , Kazimi Mujid S.
Published: November 2008

A one-dimensional single-channel thermal-hydraulics model has been developed to investigate possible occurrence of density-wave instability in two U. S. Gen-IV reactors cooled by supercritical fluids, i. e., the Supercritical Water-cooled Reactor (SCWR) and Gas-cooled Fast Reactor (GFR). Water density in the SCWR core changes from 780 kg/m3, to 90 kg/m3, whereas the density of supercritical carbon-dioxide in the reference GFR changes from 155 kg/m3 to around 110 kg/m3. The standard frequency domain approach with a decay ratio of induced velocity amplitude of 0.5 has been used to determine the onset of flow instability. With suitable inlet orificing, the hot channel of SCWR has been found to be stable. Sensitivity studies show that the hot channel decay ratio reaches the critical value of 0.5 when either the reactor power is raised to 118% of full power or the core flow rate is reduced to 86% of nominal flow rate. System pressure has only a moderate effect. Detailed 3-D studies, preferably with neutronic feedback, should be carried out for the SCWR design because of its sensitivity to various important parameters. The GFR reference design has been found to be very stable since the density change in the GFR core is rather small compared to that in the SCWR design.
Computer Code Applicability Assessment for the Advanced CANDU Reactor
Authors: Wren D. J. , Popov N. , Langman V. J. , Snell V. G.
Published: November 2008

AECL Technologies (AECLT), the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced CANDU® Reactor (ACR™)* with the United States Nuclear Regulatory Commission (USNRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990’s. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses.
A Method to Identify Dependencies Between Organizational Factors Using Statistical Independence Test
Authors: Kim Yoonik , Kim Chansoo , Jae Moosung , Chung Chang Hyun , Jung Ji Hwan
Published: November 2008

A considerable number of studies on organizational factors in nuclear power plants have been made especially in recent years, most of which have assumed organizational factors to be independent. However, since organizational factors characterize the organization in terms of safety and efficiency etc. and there would be some factors that have close relations between them. Therefore, from whatever point of view, if we want to identify the characteristics of an organization, the dependence relationships should be considered to get an accurate result. In this study the organization of a reference nuclear power plant in Korea was analyzed for the trip cases of that plant using 20 organizational factors that Jacobs and Haber had suggested. By utilizing the results of the analysis, a method to identify the dependence relationships between organizational factors is presented. The statistical independence test for the analysis result of the trip cases is adopted to reveal dependencies. This method is geared to the needs to utilize many kinds of data that has been obtained as the operating years of nuclear power plants increase, and more reliable dependence relations may be obtained by using these abundant data.
Development and Verification of Dynamics Code for Molten Salt Reactors
Authors: Krepel Jiri , Grundmann Ulrich , Rohde Ulrich
Published: November 2008

To perform transient analysis for Molten Salt Reactors (MSR), the reactor dynamics code DYN3D developed in FZR was modified for MSR applications. The MSR as a liquid fuel system can serve as a thorium breeder and also as an actinide burner. The specifics of the reactor dynamics of MSR consist in the fact, that there is direct influence of the fuel velocity to the reactivity, which is caused by the delayed neutrons precursors drift. This drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit. This leads to a reactivity loss due to the fuel flow acceleration or to the reactivity increase in the case of deceleration. For the first analyses, a 1D modified version DYN1D-MSR of the code has been developed. By means of the DYN1D-MSR, several transients typical for the liquid fuel system were analyzed. Transients due to the overcooling of fuel at the core inlet, due to the reactivity insertion, and the fuel pump trip have been considered. The results of all transient studies have shown that the dynamic behavior of MSR is stable when the coefficients of thermal feedback are negative. For studying space-dependent effects like e.g. local blockages of fuel channels, a 3D code version DYN3D-MSR will be developed. The nodal expansion method used in DYN3D for hexagonal fuel element geometry of VVER can be applied considering MSR design with hexagonal graphite channels.
Jet Pump Noise Analysis for BWRs
Authors: Castillo-Dura´n Rogelio , Ortiz-Villafuerte Javier , Herna´ndez-Lo´pez He´ctor , Alonso-Vargas Gustavo , Calleros-Micheland Gabriel
Published: November 2008

The use of noise analysis for detection of BWR component malfunction is a powerful tool in determining abnormal operation conditions, during the life of a nuclear power plant. Since the 80s, several nuclear reactors have reported problems related with jet pumps and recirculation loops. The NRC, in the IE Bulletin 80-07, recommended performing periodic monitoring to individual pressure drop jet pumps, to prevent structural failure. In this work, noise analysis methods are used for detection of jet pumps abnormal operation conditions in a BWR. Power signals obtained from the backup process computer of a BWR are analyzed with a home-developed software, called NOISE, for noise diagnostic of power signals. The computer program takes individual signals from the tabular report of the process computer. The normalized power spectral density (NPSD) is then obtained, using a Prime Factor Algorithm to calculate the Fast Fourier Transform. The NPSD of the jet pumps pressure drop, of Unit 2 of the Laguna Verde Nuclear Power Plant, showed a noticeable change in jet pump 6 during 2003, considering the period from the startup test to operation during 2003. This abnormal condition was due to that the jet pump throat was partially blocked. The noise analysis methodology is shown to be a useful tool for malfunction detection, and could be applied to create a data bank for monitoring the dynamic behavior of BWR jet pumps.
Numerical Simulation of Surface Tension Driven Flows and Boron Concentrations Around a Vapor Bubble During Sub-Cooled Boiling in Boric Acid Solution
Authors: Rao Qinyang , Jones Barclay G.
Published: November 2008

In this study, the previously developed hemispherical bubble model has been extended to a full spherical bubble, in which the bubble contact angle can range from 0° to 90°. The surface tension driven flow and boron concentration distributions around the bubble under sub-cooled boiling are simulated with constant wall temperature. Bubble contact angle and sub-cooling level are studied on their effects on boron concentration distribution. Boron precipitation under these circumstances is discussed. This is of possible concern in precipitation of boron from the solution at the triple interface during sub-cooled boiling. The deposition of porous compounds on fuel cladding provides one of the possible hideout mechanisms for Boron, which may eventually result in Axial Offset Anomaly (AOA) in PWRs.
A Simplified Model for Dynamic Simulation of the Primary Circuit of Pressurized Water Reators
Authors: Su Jian , Pontual Thiago
Published: November 2008

In this paper, we present a mathematical model for the simulation of transient neutronic and thermohydraulic behaviour of the PWR core and the primary circuit of a nuclear power plant. The point kinetics equations are used to model the core neutronics and the improved lumped parameter formulation proposed by Regis et al. (2000) and Su and Cotta (2001) is used to model the fuel dynamics. A two-region model is used for the pressurizer and for the primary-side of the steam generator. A pump model is implemented to obtain the flowrate variation during total or partial loss of flow. Several transient events and accidents are simulated and qualitatively agree with the results of more detailed engineering models.
Use of Markov Piping Reliability Models to Evaluate Time Dependent Frequencies of Loss of Coolant Accidents
Authors: Fleming Karl N. , Lydell Bengt O. Y.
Published: November 2008

Markov model theory has been applied to develop a method to evaluate the influence of alternate strategies for in-service inspection and leak detection on the frequency of leaks and ruptures in nuclear power plant piping systems [1–4]. This approach to quantification of pipe rupture frequency was originally based on a Bayes’ uncertainty analysis approach to derive piping system failure rates from a combination of service experience data and some simple reliability models [5–7]. More recently the Markov model approach has been used in conjunction with probabilistic fracture mechanics methods in the study of flow accelerated corrosion [8]. One interesting property of the Markov model is its capability to evaluate time dependent rupture frequencies via the model hazard rate. In this paper this time dependent modeling capability is used to investigate the age related and time dependent frequencies of loss of coolant accident (LOCA) initiating event frequencies. A case is presented that plant age dependent LOCA frequencies should be used in lieu of other metrics commonly used in probabilistic risk assessments and in risk informed inservice inspection evaluations. Such more commonly used metrics include the assumed constant failure rate method and the lifetime average rupture probability. Both of these methods are shown to provide optimistic estimates of LOCA frequencies for plants in the latter part of their design lifetimes, which most operating plants are approaching.
Glass Melter Replacement and Melter Technology Development in the Tokai Vitrification Facility
Authors: Aoshima Atsushi , Kozaka Tetsuo , Tanaka Kazuhiko
Published: November 2008

Japan Nuclear Cycle Development Institute (JNC) has the vitrification facility (Tokai Vitrification Facility: TVF) at the Tokai Reprocessing Plant (TRP) site (Fig. 1). In the TVF, highly radioactive liquid waste from the main process of the TRP is vitrified into glass matrix. On March of 2002, the damage on one electrode of the 1st melter occurred, so JNC decided to exchange the melter to the new one in two years. The design of the second melter to be replaced is modified to have more effective discharge ability of the platinum group metal compounds which are suspended in the molten glass matrix. The exchange project has four main steps: (1) fabrication of the new melter, (2) removal the failed melter, (3) installation of the new melter in the TVF and (4) dismantling the failed melter. As of April 2004, fabrication of the new melter and its cold operation test have been completed and the in-cell remote installation has been initiated. The installation will be finished until the end of August 2004 and the TVF operation will restart on October. For dismantling of the melter, JNC planned to cut structural materials by the YAG laser system and has progressed the device design and fabrication. The melter dismantling will be started in 2005 and it also includes detail investigation of the failed electrode. In the TVF, JNC have started other new activities for development of future melter technologies such as (1) design study of advanced melter, which has less environmental burden (2) glass volume reduction.
CAREM X INPRO Case Study
Authors: Florido Pablo C. , Delmastro Dari´o , Brasnarof Daniel , Azpitarte Osvaldo E.
Published: November 2008

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.
Magnetohydrodynamic Turbulent Model for LMFRs
Authors: Ohira Hiroaki , Ara Kuniaki
Published: November 2008

In order to study magnetohydrodynamic behavior in electromagnetic pumps, electromagnetic flow meters, etc. for Liquid Metal Fast Reactors (LMFRs), a large eddy simulation method using an artificial wall boundary condition was developed. In this study, Spalding’s law of the wall and the eddy viscosity for uniform magnetic fields, which was proposed by Shimomura, was applied to Finite Element Method of Generalized Simplified Marker and Cell (GSMAC-FEM). We calculated MHD channel flow in various element sizes on the conditions of Hartmann numbers of 0, 52.5 and 125, whose Reynolds numbers based on the average velocity were all about 29,000. These results showed the average velocity profiles were in good agreement with both the experimental results by Brouillette-Lykoudis and the detail calculation results by Shimomura, although farther calculations were needed to verify the turbulence intensities.
Burnup and Temperature Effects on CANDLE Burnup of Block-Type High Temperature Gas Cooled Reactor
Authors: Ohoka Yasunori , Sekimoto Hiroshi
Published: November 2008

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top or from top to bottom of the core and without any change in their shapes. It can be applied easily to the block-type high temperature gas cooled reactor using an appropriate burnable poison mixed with uranium oxide fuel. In the present study, the burnup distribution and the temperature distribution in the core are investigated and their effects on the CANDLE burnup core characteristics are studied. In this study, the natural gadolinium is used as the burnable poison. With the fuel enrichment of 15%, the natural gadolinium concentration of 3.0% and the fuel pin pitch of 6.6cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half width of power density distribution of 1.5m for uniform group constant case at 900K. When the effect of nuclide change by burnup is considered, the burning region speed becomes 25cm/year and the axial half-width of power density distribution becomes 1.25m. When the temperature distributions effect is considered, the effects on the core characteristics are smaller than the burnup distribution effect. The maximum fuel temperature of the parallel flow case is higher than the counter flow case.
Fiber-Optic Transmitter for Nuclear Power Plants
Authors: Higashi Junichi , Murakawa Shinichi
Published: November 2008

A promising Fiber-Optic Differential Pressure (DP) Transmitter is under development in Flexible Maintenance System (FMS) Projects that supported by Ministry of Economic, Trade, and Industries of Japan. The object of FMS projects is to improve maintenance works at nuclear power plants with latest technology. The new DP Transmitter uses optic-fiber technology of Extrinsic Fabry-Perot Sensor and Fizeau White-Light Cross-Correlator. Validation tests were performed to evaluate the tolerance of the DP transmitter in Nuclear Power Plant conditions. General requirements of PWR are accuracy (repeatability and linearity) of within +/−0.5%, pressure-proof of maximum 17.16MPa, Irradiation of 100Gy, and temperature range of 10–50 degrees centigrade at normal condition. The test results show the new DP transmitter can be expected as the next generation instrumentation in Nuclear Power Plants.
User Requirements and Criteria for Proliferation Resistance in INPRO
Authors: Choi Kun-Mo , Hurt Robert D. , Shea Thomas E. , Nishimura Richard
Published: November 2008

In designing future nuclear energy systems, it is important to consider the potential that such systems could be misused for the purpose of producing nuclear weapons. INPRO set out to provide guidance on incorporating proliferation resistance into innovative nuclear energy systems (INS). Generally two types of proliferation resistance measures are distinguished: intrinsic and extrinsic. Intrinsic features consist of technical design features that reduce the attractiveness of nuclear material for nuclear weapon program, or prevent the diversion of nuclear material or production of undeclared nuclear material for nuclear weapons. Extrinsic measures include commitments, obligations and policies of states such as the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and IAEA safeguards agreements. INPRO has produced five basic principles and five user requirements for INS. It emphasizes that INS must continue to be an unattractive means to acquire fissile material for a nuclear weapon program. It also addresses as user requirements: 1) a balanced and optimised combination of intrinsic features and extrinsic measures, 2) the development and implementation of intrinsic features, 3) an early consideration of proliferation resistance in the development of INS and 4) the utilization of intrinsic features to increase the efficiency of extrinsic measures. INPRO has also developed criteria, consisting of indicators and acceptance limits, which would be used by a state to assess how an INS satisfies those user requirements. For the first user requirement, the most important but complex one, INPRO provides a 3-layer hierarchy of indicators to assess how unattractive a specific INS would be as part of a nuclear weapon program. Attributes of nuclear material and facilities are used as indicators to assess intrinsic features. Extrinsic measures imposed on the system are also assessed. Indicators to assess defence in depth for proliferation resistance include the number and robustness of barriers, and the redundancy or complementarity of barriers. The cost of incorporating proliferation resistant features is used to assess the cost-effectiveness of any particular INS in providing proliferation resistance. The stages in the development of an INS at which proliferation resistance is considered in the process are assessed. Awareness of extrinsic measures by designers and use of intrinsic features for verification illustrate how intrinsic features facilitate extrinsic measures. An INPRO-consistent methodology to assess the proliferation resistance of an INS is still under development, with feedback expected from the case studies undertaken by Argentina, India, Russia and the Republic of Korea.
Analytical Modeling of a Bolted Flange Joint Subjected to Creep Relaxation
Authors: Nechache Akli , Bouzid Abdel-Hakim , Leˆ Van Ngan
Published: November 2008

Bolted flanged gasketed joints are the weak link between pressure vessel equipments including nuclear reactors. Their leakage tightness behavior is compromised due to the effect of creep of the gasket, the bolts and the flange, which relaxes the bolt load and causes a subsequent loss of the gasket contact stress. This is especially true when the joint is operating under high pressure and high temperature conditions. Apart from an acknowledgement of this affect, the ASME code does not give specific guidelines to help the design engineer in assessing this effect. The objective of this paper is to present an analytical method capable of predicting the bolt load relaxation in a gasketed joint as a result of creep of either the gasket, the bolt and the flange separately or all together. The proposed method is validated by comparison with 3D FE models of different size flanges and experimental data. A strong emphasis will be put on flange rigidity, which is the major controlling parameter of the load relaxation.
Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability
Authors: de Rocquigny Etienne , Chevalier Yoan , Turato Silvia , Meister Eric
Published: November 2008

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.
Linear and Nonlinear Stability Analysis of a Self-Pressurized, Natural Circulation, Integral Reactor
Authors: Zanocco Pablo , Delmastro Dari´o , Gime´nez Marcelo
Published: November 2008

The stability of a self-pressurized, integral reactor cooled by natural circulation is analyzed. CAREM reactor prototype is taken as reference for the present study. Because of the self-pressurization condition, the system is very close to saturation, and some boiling occurs along the hot leg - reactor core and riser -. Possible instabilities caused by this condition are analyzed in this work in the nominal pressure, with special attention to some particular issues of this design. A numerical code is developed to describe the reactor dynamics and it is briefly described. It includes the coolant with a one-dimensional scheme, steam dome and core modelling, considering the flashing phenomena, neutronic and pressure feedbacks. The code numerically solves the whole set of non-linear equations in a time-domain approach; it also includes a linearization method by numerical perturbations, then a classical frequency-domain stability analysis can be also carried out for the linearized system, by means of eigenvalues calculation. The amplification factor for the oscillation is plotted in the core power/steam dome condensation parameter-plane. A parametric study is carried out, to analyze the influence of core dynamic and the pressure feedback due to the self-pressurization. The pressure feedback has a stabilizing effect, increasing stability for smaller steam volumes. The core dynamic has a stabilizing effect when the core power and the steam dome condensation are low, and relatively destabilizing when they are higher. The stability boundary is determined, for several cases. The amplification factor is enlarged in the region where the flashing phenomenon occurs. The dynamic non-linear effects are studied by means of a time-domain approach, for selected conditions in growing and damped oscillations. The limit-cycle for the growing oscillations is determined, and the main non-linear sources are studied: the amplitude is restricted by strong non-linear effects, which appear when the boiling boundary crosses the core-riser limit, and more than one boiling point coexists at the same time, bounding the amplitude of oscillation to values that would remain un-noticed in case of being present in the reactor.
Hard Spectrum STFR (Spray Type Fast Reactor) Cooled by Water at 7.5 MPa
Authors: Kitamoto Asashi , Ohoka Yasunori
Published: November 2008

New concept of fast reactor (FR), i.e. the STFR (Spray Type FR), is proposed here to perform high burn-up of UO2 fuel or MOX fuel by the use of BWR technology, and to improve the backend process of discharged fuel. STFR can be realized by some important changes of BWR system, at 7.50MPa. In case of STFR, heat produced in the core is removed by the evaporation of sprayed hot water jetted to fuel with cross flow at 7.50MPa, and two phases (liquid and vapor) of coolant at high void ratio goes down to the bottom of PV (pressure vessel). This is an improved concept of BWR, which can be regarded as a breakthrough of FBR. This concept has not been listed in GEN IV. Future performance of STFR are as follows, (1) STFR can increase the fraction of direct fission of 238U with neutron reaction of higher energy than 1MeV, (2) STFR can burn the nuclear fuel to the higher burn-up (80 to 200GWd/Mg-HM) compared with BWR fuel burn-up. (3) Higher burn-up of fuel will reduce the frequency of reprocessing, so STFR can reduce the reprocessing cost per power production. (4) STFR can reduce the remains of Pu and MA (Minor Actinides: Np, Am, Cm etc.) in discharged fuel.
The State-of-the-Art of the LTAN Method in the Solution of the Transport Equation in a Slab
Authors: Cardona Augusto Vieira , Panta Pazos Rube´n , de Vilhena Marco Tu´llio
Published: November 2008

In this work we report the state-of-the-art of the LTAN method, reporting the derivation of the LTAN homogeneous and particular solution in a slab for large N (N < 650), as well its accelerated version, which reduces drastically the computational time. We also sketch the mathematical background for the error bound estimate and convergence analysis of the LTAN method. Finally, we display numerical results and comments.
Applications of the Master Curve Approach to Irradiated Steels
Authors: Ballesteros A. , Bros J. , Brumovsky M.
Published: November 2008

The Master Curve approach is being introduced in the evaluation of the fracture toughness of low-alloy ferritic steels used in reactor pressure vessels. The Master Curve provides significant technical enhancements compared with the methodology currently incorporated in the ASME code. The fundamental understanding of Master Curve has been achieved, but application issues are emerging facing the integrity evaluation of reactor pressure vessels. The current situation of the Master Curve methodology is reviewed in this paper, including the last developments for WWER reactors. The open issues are discussed, and the preliminary results from investigations being carried out in Spain are presented.
Mitigation of a Core Meltdown Scenario by Means of a Core Catcher Located Inside the Reactor Pressure Vessel
Authors: Aquaro D. , Zaccari N.
Published: November 2008

This paper describes an original solution of core catcher to managing the in vessel retention of the Corium in the accidental event of the core meltdown. The solution envisaged intends to verify the possibility of managing the accidental event within the pressure vessel, ensuring that the CORIUM is confined and cooled. The core catcher, elaborated at the DIMNP, is made of a ceramic pebbel bed (Alumina Al2O3) contained in a metallic or Ceramic Matrix Composite (CMC) structure. The paper illustrates a theoretical model to simulate the thermal-mechanical behaviour of the pebble beds under extremely high loads, developed by the authors. This model has been used to design the core catcher and to determine the effective conductivity and the effective stiffness of the pebble bed. These values have been used in order to implement a numerical model of the core catcher. The results of the thermal and mechanical coupled simulation have permitted to determine the maximum time that the core catcher could resist and the mechanical resistance of the core catcher in the case of RPV external or internal cooling. The preliminary analyses performed have emphasised the good performance of pebble bed core catcher in order to mitigate the envisaged severe accident.