bluuebunny/crossref_metadata_2025 huggingface dataset:
Properties, Weldability, and Applications of Modern, Wrought, Heat-Resistant Alloys for Aerospace and Power Generation Industries
Authors: Rowe M. D. , Ishwar V. R. , Klarstrom D. L.
Published: November 2008

Alloy selection and alloy design both require consideration of an array of material attributes, including in-service properties, weldability, and fabricability. Critical properties of modern heat-resistant alloys for gas turbine applications include high temperature strength, thermal stability, oxidation resistance, and fatigue resistance. In this paper, the properties of twelve solid-solution-strengthened and six age-hardenable heat-resistant alloys are compared. Weldability is an important attribute, and can be a major limiting factor in the use of certain alloys. Weldability test methods are discussed and the resistance of alloys to solidification cracking and strain-age cracking is compared. The use of weldability testing in the development of modern heat-resistant alloys is discussed with several examples cited. Finally, alloy selection for gas turbine components is outlined, taking into account both alloy properties and fabricability.
Application of Continuous Combustion Dynamics Monitoring on Large Industrial Gas Turbines
Authors: Sewell Jesse , Sobieski Pete , Beers Craig
Published: November 2008

This paper presents results from continuous monitoring of combustion dynamics and its application in determining the health of combustion system hardware. A number of 180 MW class large industrial gas turbines operating on natural gas have been instrumented with a continuous Combustor Dynamics Monitoring (CDM) for each combustion chamber. Tuning considerations for emissions, stability and extended combustion parts life is discussed. Physical inspection of engine hardware is correlated with spectral and supervisory engine data to determine signatures in the combustion chambers that represent damaged or failing hardware. This methodology can also be used to identify potentially harmful operational profiles. Spectral examples of a pilot nozzle and transition piece failures are presented. Monitoring and recognizing the differences in spectral signatures associated with stability and component failures suggests better understanding of combustion dynamics contribution to combustor parts wear and reduction in downstream damage.
OH PLIF and Soot Volume Fraction Imaging in the Reaction Zone of a Liquid-Fueled Model Gas-Turbine Combustor
Authors: Meyer Terrence R. , Roy Sukesh , Gogineni Sivaram P. , Belovich Vincent M. , Corporan Edwin , Gord James R.
Published: November 2008

Simultaneous measurements of OH planar laser-induced fluorescence (PLIF) and laser-induced incandescence (LII) are used to characterize the flame structure and soot formation process in the reaction zone of a swirl-stabilized, JP-8-fueled model gas-turbine combustor. Studies are performed at atmospheric pressure with heated inlet air and primary-zone equivalence ratios from 0.55 to 1.3. At low equivalence ratios (φ < 0.9), large-scale structures entrain rich pockets of fuel and air deep into the flame layer; at higher equivalence ratios, these pockets grow in size and prominence, escape the OH-oxidation zone, and serve as sites for soot inception. Data are used to visualize soot development as well as to qualitatively track changes in overall soot volume fraction as a function of fuel-air ratio and fuel composition. The utility of the OH-PLIF and LII measurement system for test rig diagnostics is further demonstrated for the study of soot-mitigating additives.
Modelling a New Three-Pad Active Bearing
Authors: Krodkiewski Janusz M. , Davies Gregory J.
Published: November 2008

This paper describes investigations into a new type of active bearing to be implemented in the field of rotating machinery. Active bearings are based on the concept that journal oil flow can be modified during operation by active devices. Here, the concepts of the flexible pads and the oil-filled chambers that control their deflection are used. Three active pads are positioned around the journal with three oil-filled chambers positioned behind them. One of them, the load-bearing pad located at the bottom of the bearing, acts as a passive device and is equipped with a very thin film chamber, which acts as a source of damping. Such a damper was found in previous work to be effective in dissipating energy. Here, in a departure from previous work, two additional small pads with deep oil-filled chambers have been added in order to allow control theory to be implemented. They are located in the upper part of the active bearing. A non-linear system model is developed for the rotor-bearing system that includes the described active bearing. The flow inside the upper chambers that control motion of the active pads was neglected due to their large volume. It results in a uniform pressure distribution along the upper pads. The pressure distribution within the damper oil film (inside the lower chamber) and the journal oil film was modeled with the aid of the Reynolds equation. They were solved by means of the finite difference method and Gauss-Seidel technique. The same mesh used for solution of the Reynolds equation was used for the division of the flexible pads into the finite elements. The same approach was adopted for the modelling of the dynamic properties of the rotor. The mass and stiffness matrices for the pads and rotor were condensed down to 12 master generalized coordinates using Guyan condensation. The obtained system of matrix equations was converted to a system of first order equations and solved via the Runge-Kutta method. Some results of the numerical testing of the mathematical model developed are provided.
Turbine Blade Surface Deterioration by Erosion
Authors: Hamed Awatef A. , Tabakoff Widen , Rivir Richard B. , Das Kaushik , Arora Puneet
Published: November 2008

This paper presents the results of a combined experimental and computational research program to investigate turbine vane and blade material surface deterioration caused by solid particle impacts. Tests are conducted in the erosion wind tunnel for coated and uncoated blade materials at various impact conditions. Surface roughness measurements obtained prior and subsequent to the erosion tests are used to characterize the change in roughness caused by erosion. Numerical simulations for the three dimensional flow field and particle trajectories through a low pressure gas turbine are employed to determine the particle impact conditions with stator vanes and rotor blades using experimentally-based particle restitution models. Experimental results are presented for the measured blade material/coating erosion and surface roughness. The measurements indicate that both erosion and surface roughness increase with impact angle and particle size. Computational results are presented for the particle trajectories though the first stage of a low-pressure turbine of a high bypass turbofan engine. The trajectories indicate that the particles impact the vane pressure surface and the aft part of the suction surface. The impacts reduce the particle momentum through the stator but increase it through the rotor. Vane and blade surface erosion patterns are predicted based on the computed trajectories and the experimentally measured blade coating erosion characteristics.
Influence of Fuel Chemical Composition on Particulate Matter Emissions of a Turbine Engine
Authors: Corporan Edwin , Monroig Orvin , Wagner Matthew , Dewitt Matthew J.
Published: November 2008

The effects of fuel chemical composition on particulate matter (PM) emissions of a T63 engine were investigated. Fuels with different aromatic, cycloparaffin (naphthene), iso-paraffin and normal paraffin levels were evaluated in the turboshaft engine and compared to the performance of a typical JP-8 fuel. The fuels studied include: a semi-synthetic jet fuel, two high naphthenic experimental fuels, three Exxon solvents (Isopar M, Isopar H and Norpar-13) and methylcyclohexane. The effect of blending solvents in JP-8 on PM emissions was also assessed. Commercially available particulate instruments were used to measure particle number density, particulate mass concentration and particle size distribution. Results showed a general trend of higher particulate concentrations and larger diameter soot particles with decreasing fuel hydrogen-to-carbon (H/C) ratio or increasing aromatic content. However, for several fuels with approximately the same H/C ratio, significant differences in PND and calculated mass were observed. Furthermore, blends of JP-8 with solvents of similar H/C ratio but varying chemical composition produced significantly different particulate emissions. These findings demonstrate that particulate emissions from hydrocarbon-fueled combustion processes are not solely a function of the H/C ratio or aromatic concentration of the fuel, but that other properties or constituents also impact soot emissions. Chemical and physical properties of the fuels and their potential effect on particulate emissions are discussed. These findings provide insight into fuel properties that impact PM emissions, which may aid in the development of fuel additives to reduce particulate emissions from turbine engine combustors.
The Development of Innovative Methods for Erosion Testing a Russian Coating on GE T64 Gas Turbine Engine Compressor Blades
Authors: Klein Mark , Simpson Gray
Published: November 2008

The US Navy became aware of a Russian erosion-resistant compressor coating in 1991. The coating was being applied to compressor airfoils on the TV2-117 engine (MI-8 helicopter). Performance data obtained from the Russian coating company showed greatly improved erosion resistance, leading to better engine performance and longer time-on-wing. A proposal was sent to the Foreign Comparative Testing (FCT) office to evaluate the coating for US Navy engines operating in sandy environments. The project was approved and funded in 1997. It was determined that the high-speed erosion chamber at the University of Cincinnati would provide the optimum testing environment for the program. The test articles consisted of a bare blade and a coated blade. Testing conditions required 10 micron silica sand impinging upon the blades at 700 feet per second (fps). Data was collected until the bare blade reached the chord reject criteria. Testing continued until the coated blade reached failure as well. It was found that the coating improved the overall blade life by 300–500%. A method was developed to measure the eroded blades at certain points to determine the real or “effective” chord at any point during the testing. This project demonstrated the need for innovative gauging techniques in order to understand the actual effects of erosion on compressor blades.
Development of an Advanced Microturbine System Using Humid Air Turbine Cycle
Authors: Dodo Satoshi , Nakano Susumu , Inoue Tomoaki , Ichinose Masaya , Yagi Manabu , Tsubouchi Kuniyoshi , Yamaguchi Kazuyuki , Hayasaka Yasushi
Published: November 2008

A prototype machine for a next generation microturbine system applying a simple humid air turbine system (design target of electrical output: 150 kW, electrical efficiency: 35% LHV) was developed for its laboratory evaluation. A low NOx combustor which applied a lean-lean zone combustion concept and water lubricated bearings were developed for the prototype machine. Operation using two water lines for the humid air turbine (HAT) was proposed as an effective way to obtain rated electric output to ambient temperature of 40 deg C. Tests for the main components were done successfully. Motoring tests, full speed test with no load, 50% load and 70% load tests as preliminary tests for rated load tests were also carried out successfully. Low NOx emission of 7.6 ppm and high efficiency of 95.6% for the power conversion system were achieved in the partial load tests. At the first rated load test without HAT and Water atomizing inlet air cooling (WAC) that followed those partial load tests, 150.3 kW electric output with electrical efficiency of 32% was obtained.
A Multiple Decision Support Metrics Method for Effective Risk-Informed Asset Management
Authors: Liming James K. , Salter James E.
Published: November 2008

The objective of this paper is to provide electric utilities with a concept for developing and applying effective decision support metrics via integrated risk-informed asset management (RIAM) programs for power stations and generating companies. RIAM is a process by which analysts review historical performance and develop predictive logic models and data analyses to predict critical decision support figures-of-merit (or metrics) for generating station managers and electric utility company executives. These metrics include, but are not limited to, the following: profitability, net benefit, benefit-to-cost ratio, projected return on investment, projected revenue, projected costs, asset value, safety (catastrophic facility damage frequency and consequences, etc.), power production availability (capacity factor, etc.), efficiency (heat rate), and others. RIAM applies probabilistic safety assessment (PSA) techniques and generates predictions probabilistically so that metrics information can be supplied to managers in terms of probability distributions as well as point estimates. This enables the managers to apply the concept of “confidence levels” in their critical decision-making processes.
Radiation Safety Management for Manned Underwater Operations at Nuclear Facilities
Authors: Vallance C. A. , Jackson Matthew , Wilkins Kyle
Published: November 2008

Commercial divers are frequently utilized to perform critical work at commercial nuclear power plants, fuel reprocessing plants, waste storage facilities, and research institutions. Various types of tasks can be performed underwater including inspection, welding, mechanical modifications, repairs, coating, and work associated with decommissioning. Work may take place in areas such as the reactor vessel, equipment pool, spent fuel pool, and suppression chamber. Because water provides excellent shielding, divers are able to work in areas where radiological exposures would be prohibitive or even fatal to workers in dry conditions. To perform such work safely while minimizing exposure requires special equipment and techniques. This paper will provide an overview of the procedures and equipment necessary to perform work in radiation areas while meeting ALARA (As Low As Reasonably Achievable) goals. Careful planning and execution are necessary and require close cooperation with site personal. Dive teams must coordinate with operations, maintenance, radiation protection, radwaste disposal, chemistry, and other departments. Examples of representative projects are used to illustrate key points.
Optimization of the Westinghouse Sulfur Process for Hydrogen Generation and the Interface With an HTGR
Authors: Lahoda Edward J. , Goossen John E. , Matzie Regis A. , Mazzoccoli Jason P.
Published: November 2008

The Westinghouse Sulfur Process is a member of the sulfur family of hydrogen generating thermochemical cycles being considered by the DOE for coupling with an HTGR. It has been proven experimentally and utilizes mature technology. Westinghouse has identified process improvements that increase the efficiency and reduce materials and HTGR-Hydrogen Process integration issues. Increasing the hydrogen plant operating pressure improves the thermal efficiency of the Westinghouse Sulfur Process from 40% to ∼55% (LHV basis) at H2SO4 concentrations of 40% to 80% by weight. The use of a directly heated decomposition reactor in conjunction with an HTGR allows higher decomposition reactor operating temperatures. This increases the per-pass percent conversion and increases overall efficiency. Other advantages are total separation of the coolant and process streams, the maturity of this technology (it has been used for over a century in the steel industry at temperatures up to 1,200°C), and the ready availability of materials.
Plant Modifications for Mixed Oxide Fuel Use in U.S. Reactors
Authors: Browder Richard
Published: November 2008

Duke Power’s McGuire Nuclear Station and Catawba Nuclear Station have been selected to aid in the disposal of weapons grade plutonium. In order for McGuire and Catawba to transition to a 40% MOX core, several plant modifications are required. Among these are: 1) transition to enriched soluble boron, 2) security enhancements due to presence of strategic special nuclear materials, 3) transportation interface modifications, and 4) reactor protection and control modifications. This paper describes the modifications that will enable the timely disposition of surplus weapons grade plutonium in the McGuire and Catawba reactors.
One-Dimensional Drift-Flux Model and Constitutive Equations for Relative Motion Between Phases in Various Two-Phase Flow Regimes at Microgravity Conditions
Authors: Hibiki Takashi , Takamasa Tomoji , Ishii Mamoru
Published: November 2008

In view of the practical importance of the drift-flux model for two-phase flow analyses at microgravity conditions, the constitutive equations for distribution parameter and drift velocity have been developed for various two-phase flow regimes at microgravity conditions. A comparison of the model with various experimental data over various flow regimes and a wide range of flow parameters taken at microgravity conditions shows a satisfactory agreement. The newly developed drift-flux model has been applied to reduced gravity conditions such as 1.62 and 3.71 cm/s2, which correspond to the Lunar and Martian surface gravities, respectively, and the effect of the gravity on the void fraction in two-phase flow systems has been discussed.
The Adventitious-Pin-Failure Study Under a Slow Power Ramp
Authors: Fukano Yoshitaka , Charpenel Jean
Published: November 2008

In a fast breeder reactor, a slow power ramp accident could lead to a local melting of the fuel depending on design and assumptions. If we assume cladding failure in addition to the fuel melting, possibility of fuel ejection from the pin and subsequent consequences of such fuel ejection become important for safety consideration. In order to answer this question, the RB1 and RB2 tests in the CABRI-RAFT program (from 1996 to 2002) were performed under the collaboration between the French ‘Institut de Radioprotection et de Suˆrete´ Nucle´aire’ (IRSN) and JNC. These tests were aiming at a study on impact of fuel pin failure under an overpower condition corresponding to a control-rod withdrawal-type incident leading to fuel melting. Using a special technique, combination of through-cladding failure and fuel melting was realized. In the RB1 test, fuel ejection was prevented under a limited fuel melting condition. On the other hand, significant fuel melting was applied in the RB2 test so as to get the fuel ejection, thereby obtaining information on the fuel ejection behavior. Interpretation for these tests through the detailed experimental data evaluation and the analysis with PAPAS-2S and PHYSURAC codes are performed in this study. Through this study, it is indicated that molten fuel ejection can be prevented with the low smear density fuel as far as the fuel melting is not large for a slit-type cladding defect. Fuel ejection becomes possible in the case of significant fuel melting with a very thin solid fuel shell surrounding the molten fuel cavity. However, the rapidness of the fuel ejection with the low smear density fuel is less pronounced when compared with that of the high smear density fuel. It should be noted that the coolant channel basically kept its wetness by liquid sodium in the RB2 test in the early part of post fuel ejection phase. This fact is encouraging us concerning coolability of the ejected fuel. It is also confirmed that there is considerable DN (Delayed Neutron) -precursor release into the coolant flow already before fuel ejection. This result is useful for evaluation of anomaly detection with DN signal observation.
Review and Perspective on Approaches of Evaluating Cleavage Fracture Toughness of Pressure Vessel Steels
Authors: Zhang X. Z. , Knott J. F.
Published: November 2008

Although efforts have been made to establish cleavage fracture model in the past 30 years or so, uncertainties remain in predicting the cleavage fracture toughness behavior of pressure vessel steels in the transition region. Confusions may be raised in cases such as that there are still outliers while ASME fracture toughness curve indexed to the reference temperature is generally considered as very conservative. The advantage and potentials of using Finite Element method to predict the cleavage fracture behavior of irradiated materials are obvious. However, its effectiveness depends on the rightness of the physical model and criteria established. In this paper, approaches of describing cleavage fracture toughness of pressure vessel steels in the transition region are discussed and compared. The need for focusing on the physical fundamental and criteria of cleavage fracture is addressed.
The Role of Two-Phase Coolant in Moderating Fretting in Nuclear Steam Generators
Authors: Dyke John M.
Published: November 2008

This paper expands the principal of coolant-cushioning in Nuclear Steam Generators whereby the two-phase coolant, especially the bubble film on the tube surface, moderates the vibration of coolant tubes against their supports. This thesis was advanced by the author in JSME ICONE-11 paper #36443 “Fretting in Nuclear Steam Generators – A New Approach”. The current paper addresses tube bundle and AVB geometry issues; examines the tube bundle-coolant-AVB interfaces and examines implications for recirculation flow, AVB design and boiler size. The need for research to confirm empirical findings is reiterated.
A Study on the Vitrification Tests of the Low Level Radioactive Wastes Using Plasma Arc Melter System
Authors: Yoon Boo Ho , Cho Jae Hak , Lee Sang Chul , Kang Dong Woo , Choi Yong Joon , Moon Kil Ho , Um Dallsun
Published: November 2008

For the research on the vitrification of the low-level radioactive wastes (LLRW) produced in nuclear power plants, one pilot plant with plasma arc melter system was built and several tests were done on it. Some surrogate wastes, which were spiked with several materials and were made very similar to the real LLRW, were used for these tests. For the vitrification of the surrogate wastes, the surrogate wastes were classified into the combustible, the non-combustible and the resin. Then each waste was spiked with special materials and was melted in separate. Off-gases produced in each test were picked up and analyzed. Real radioactive materials cesium (Cs) and cobalt (Co) were spiked in each wastes. Data gained from the final glass formulation were follows. Glass density is 2.42 ∼ 2.95(g/cm3), the compressive strength is 30 ∼ 175 Mpa, micro hardness is 5.5 ∼ 5.8 Gpa. The leaching ratio for Co is 1.27×10−4 ∼ 1.08×10−3 (10mL/g) and that for Cs is 2.46×10−3 ∼ 3.23×10−2 (10mL/g). The leaching speed for Co is 4.14×10−7 ∼ 5.53×10−6 (g/m2) and that for Cs is 4.58×10−5 ∼ 3.87×10−4 (g/m2). In off-gas, dioxin & furan is 0.016 mano gram on the average, CO is about 20 ppm, NO2 is about 15 ppm and SO2 is about 15 ppm.
Corrosion Behavior of Steels in Lead-Bismuth Eutectic: In Stagnant LBE Test at Low Oxygen Partial Pressure
Authors: Furukawa Tomohiro , Yoshida Eiichi , Aoto Kazumi
Published: November 2008

Corrosion test of three chromium-molybdenum steels — 2.25Cr-1Mo, Mod.9Cr-1Mo and 12Cr-steel (ASME P122) — has been performed in stagnant lead–bismuth eutectic (LBE) at low oxygen partial pressure at 550 °C up to 2,000 hours. After the test, the specimens were metallurgically examined using a scanning electron microscopy. All steels had good resistance to LBE under this test condition, and the difference of the effect of chromium concentration in the steels on corrosion behavior was not observed. Corrosion depth of the steels was a few microns, and chromium solution into LBE was observed on the surface. Very thin chromium oxide layer was partly observed on the surfaces of all steels. At the area, no corrosion was occurred. The behavior was the same as the results of martensitic steels tested in stagnant LBE containing 10−7 wt% to 10−8 wt% of oxygen.
Development of Passive Cooling System Models for the Assessment of the Long Term Cooling Capability in the Liquid Metal Reactor
Authors: Chang W. P. , Ha K. S. , Jeong H. Y. , Heo S. , Lee Y. B.
Published: November 2008

This study has been carried out to assess the decay heat removal capability of the passive safety systems adopted in a conceptual design of the 600 MW(e), sodium cooled, metallic fuel loaded KALIMER. The applicability of the PVCS, which used to be the only passive safety system for KALIMER-150, is limited to a reactor capacity of 1,000 MW(t) or less. Another passive loop, PDRC, is conceptualized in order to overcome the limit as the KALIMER capacity scales up from the current 150 MW(e) to 600 MW(e). The safety analysis computer code, SSC-K, currently used for KALIMER is not capable of simulating such passive systems. With this concern, the PVCS and PDRC models are developed and they are coupled with the SSC-K for a long-term cooling assessment. The present paper thus presents the analysis results of the ULOHS using these models along with their brief introductions. The primary concern of the analyses is focused on the inherent safety as well as the system’s integrity for 72 hours without any operator action during the event.
Simulation of Hydrogen Distribution in Containment Studies Facility: A Parametric Study Using the CFD Code FDS
Authors: Sharma Pavan K. , Markandeya S. G. , Ghosh A. K. , Kushwaha H. S.
Published: November 2008

Hydrogen transport behavior in the containment of NPPs is an extremely complex phenomenon and must be properly understood for the assessment of the overall plant safety under accident conditions. A case study has been carried out to predict the hydrogen concentration profile in the multi-floor multi-compartment geometry of Containment Studies Facility (CSF), to understand the complex hydrogen transport behavior in general and the influence of the location and direction of injection of hydrogen, in particular. A computer code FDS has been used for conducting these studies. The results indicate a strong possibility of stratification of the lighter gas within as well as across the compartments during both the injection and post injection phases. After the momentum effects due to the convective currents die down, stratification was noted to be more or less stabilized. Molecular diffusion was seen to have insignificant diffusion.
The Application of Advanced Computational Methods to the Modeling of PBMR Source Terms
Authors: Mkhosi Margaret Msongi , Denning Richard , Nakamura Shoichiro
Published: November 2008

The pebble bed modular reactor (PBMR) is a high temperature gas-cooled reactor which uses helium gas as a coolant. The PBMR design relies on the excellent heat transfer properties of graphite and a fuel design that is inherently resistant to the release of the radioactive material up to high temperatures. The safety characteristics of the PBMR concept are excellent. However, a very strong safety case will have to be made if a new generation of reactors is to be successfully introduced to a concerned public. Until recent developments in computational fluid dynamics methods, computer speed, and data storage, the coupled thermal-hydraulic, chemical, and mass transport phenomena could not be treated in an integrated analysis. This paper addresses one aspect of the interplay between the details of fluid flow and aerosol transport within the complex geometry of the pebble bed core. A very large quantity of graphite dust is produced by the interaction among the pebbles. The potential for the deposition of radionuclides on the surface of dust particles and their subsequent transport as aerosols is substantial. This effort focuses on the inertial deposition of these aerosols within the pebble bed. Inertial deposition in the low Reynolds number regime of laminar flow in pebble beds has been explored previously, but with less powerful computational techniques. Some experimental data are also available in this regime. No analyses or experimental data are available in the high Reynolds number turbulent regime in which the PBMR operates. This paper describes results of analyses of inertial deposition obtained with the FLUENT computational fluid dynamics code. The objective of the analysis is to obtain an expression for deposition within an asymptotic unit cell, removed from the boundary conditions at the entrance to the array. The results of analyses performed at different velocities and fluid densities in the turbulent regime were correlated against a modified Stokes number. The deposition correlation is well represented by the integral form of the normal distribution. Deposition for the time-averaged flow was found to be insensitive to the flow model. In the laminar regime, FLUENT results were found to be in agreement with earlier published results and experimental data. The stochastic behavior of eddies was also simulated within FLUENT using the k–ε model. Eddy-enhanced deposition results in greater deposition at all aerosol sizes in comparison with the time-averaged results, with significant deposition of aerosols predicted for small aerosol sizes. However, it is likely that these results are quite sensitive to the modeling of turbulence and they must be considered preliminary.
Stochastic Maintenance Optimization at Candu Power Plants
Authors: Doyle E. Kevin , Duschesne T. , Lee C. G. , Cho D. I.
Published: November 2008

The use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems as previously reported at ICONE 6 in New Orleans (1996). Further refinement of the station maintenance strategy was evaluated via the applicability of statistical analysis of historical failure data. The viability of stochastic methods in Candu maintenance was illustrated at ICONE 10 in Washington DC (2002). The next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principal impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring “baby boom generation”. Initial indications have been quite positive. The current reality of global competition necessitates the pursuit of all financial optimizers. The next construction phase in the power generation industry will soon begin on a worldwide basis. With the relatively high initial capital cost of new nuclear generation all possible avenues of financial optimization must be evaluated and implemented.
Steady State Analysis of Molten Salt Reactor in Consideration of the Effect of Fuel Salt Flow
Authors: Yamamoto Takahisa , Mitachi Koshi , Suzuki Takashi
Published: November 2008

The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows the inside of the reactor accompanying nuclear fission reaction. In the previous study, the authors had developed numerical model to simulate the effects of the fuel salt flow on the reactor characteristics. This paper applies the model to the steady state analysis of the small MSR system and estimates the effects of the fuel flow. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The following results are obtained: (1) the fuel salt flow affects the distributions of the delayed neutron precursors, especially long-lived one, and (2) the extension of residence time in the external loop system and the rise of fuel inflow temperature slightly show negative reactivity effects, decreasing neutron multiplication factor of the small MSR system.
Dynamic Stiffness Formulation for Torsional Vibrations of Generally Restrained Thin-Walled Beams of Open Section Resting on Continuous Elastic Foundation
Authors: Kameswara Rao Chellapilla , Srinivasulu N. V.
Published: November 2008

This paper deals with the dynamic stiffness matrix formulation for the problem of torsional vibrations and stability of generally restrained doubly symmetric thinwalled beams of open section, resting on Winkler type continuous elastic foundation. An analytical method is developed, including the effects of warping and neglecting the effects of longitudinal inertia and shear deformation. The resulting transcendental frequency equation can be solved for various values of linear and rotational end restraint parameters. A computer program is developed, based on the dynamic stiffness matrix approach. The software consists of a master program to set up the dynamic stiffness matrix and to call specific subroutines to perform various system calculations. To demonstrate the influence of end restraints and warping on the torsional natural frequency, numerical results are obtained for the case of cantilever thinwalled beam of open section partially restrained against warping for various values of restraint parameter and warping parameter and the same are presented in suitable form.
Validation of Physics and Thermalhydraulics Computer Codes for Advanced CANDU Reactor Applications
Authors: Wren D. J. , Popov N. , Snell V. G.
Published: November 2008

Atomic Energy of Canada Ltd. (AECL) is developing an Advanced CANDU Reactor (ACR) that is an evolutionary advancement of the currently operating CANDU 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular CANDU concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the CANDU 6, most of the key phenomena associated with the safety analyses of ACR are common, and the CANDU industry standard toolset of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design features that have impact on the physics and thermalhydraulics codes.
Large-Break Loss-of-Coolant Accident Phenomena Identification and Ranking Table (PIRT) for the Advanced CANDU Reactor
Authors: Popov N. , Sills H. E. , Snell V. G. , Boyack B. , Langman V. J.
Published: November 2008

The Advanced CANDU Reactor (ACR™)* is an evolutionary advancement of the current CANDU 6® reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all CANDU reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table.
Analyses of Ex-Vessel Steam Explosion and Its Structural Dynamic Response for a Typical PWR Plant
Authors: Kawabata Osamu
Published: November 2008

The present paper deals with analysis of an ex-vessel steam explosion and its structural dynamic response to confirm the integrity of a concrete structure simulating the reactor cavity compartment of the four loop PWR plant. The VESUVIUS code, which has been developed in the IMPACT project of NUPEC, was used for analysis of the molten-core coolant interaction phenomenon. The steam explosion model in VESUVIUS code contains break-up of melted jet, pre-mixing, and propagation of impulsion wave in the cavity water with a multiple-dimensional model. Initial conditions were determined based on the KROTOS steam explosion experiment performed at the JRC-Ispra in Italy. In a pre-mixing phase analysis, a melted jet entering into the water and the associated breakup phenomenon were analyzed with various water depths. The result showed that a spontaneous steam explosion was not predicted in the case of 3m water depth. However, a spontaneous steam explosion condition was onset in the case of 5m water depth. The results showed that the maximum pressure load to the cavity floor reached to 17MPa in the case of 5m water depth due to pressure propagation through fine fragmentation process. The conversion efficiency of mechanical energy was estimated to be about 2.9% in the compulsive triggering case with 5m water depth. The structural dynamic response analysis of ferroconcrete around the reactor vessel was performed by the AUTODYN-2D code, which is two-dimensional impact analysis code, applying analytical results obtained with the VESUVIUS code. The maximum value of the plasticity strain of concrete was estimated about 2.4%, and therefore calculated results indicated the possibility of local damage to the concrete structure. However, the re-bar remained in elastic region, and the integrity of a concrete structure was secured.
Cable Tray Ultimate Strength Test Employing a Large Shaker Table
Authors: Komatsu Kuniyoshi , Myojin Katsunori , Fuyama Hiroyuki , Kokubo Eiji , Imai Kazuo
Published: November 2008

Ultimate behaviors of cable trays, used in nuclear plants, have not been well studied since cable trays are designed based on conservative design criteria. In this study, by employing a large shaker table, an ultimate strength test was conducted for cable trays used in nuclear power plants. This report describes the results of shaker table test. The following results were obtained: 1) In an S2* earthquake, the damping ratio was so large — more than 30% due to the rubbing of cables — that a large response was not present and the strains in the support were within the elastic limits. 2) The support was strong enough to sustain the cable trays even when the strain in the support was 20 times larger than elastic strain.
Experimental Investigation of Evaporation Behavior for Lead-Bismuth Eutectic and Its Impurity Tellurium
Authors: Ohno Shuji , Miyahara Shinya , Kurata Yuji
Published: November 2008

Evaporation behavior of lead-bismuth eutectic (LBE) and of tellurium in LBE pool is being investigated experimentally in order to obtain fundamental data on the coolant technology of an accelerator driven system and a fast reactor. The tellurium was used as a non-radioactive chemical surrogate of polonium which would generate and give radiological hazards in the above LBE cooled system. The experiments were conducted using the transpiration method in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. The LBE pool was about 500grams in mass and had a surface area of 4cm × 14cm. Two kinds of experiment were performed under the temperature condition in the range of 450 to 750°C to measure the collected vapor mass: LBE experiment without impurity and LBE-tellurium experiment. The paper presents experiment outline and several instructive results such as saturated vapor pressure of LBE, information on evaporation behavior of each element, and equilibrium partitioning behavior of tellurium from liquid to gas phase. These are also suggestive for understanding chemical behavior of polonium in LBE.
Application of ANNs in Tube CHF Prediction: Effect of Neuron Number in Hidden Layer
Authors: Han Lang , Shan Jianqiang , Zhang Bin
Published: November 2008

Prediction of the Critical Heat Flux (CHF) for upward flow of water in uniformly heated vertical round tube is studied with Artificial Neuron Networks (ANNs) method utilizing different neuron number in hidden layers. This study is based on thermal equilibrium conditions. The neuron number in hidden layers is chosen to vary from 5 to 30 with the step of 5. The effect due to the variety of the neuron number in hidden layers is analyzed. The analysis shows that the neuron number in hidden layers should be appropriate, too less will affect the prediction accuracy and too much may result in abnormal parametric trends. It is concluded that the appropriate neuron number in two hidden layers should be [15 15] in the article.
Development of Simplified Compact Containment BWR Plant
Authors: Heki H. , Nakamaru M. , Tsutagawa M. , Hiraiwa K. , Arai K. , Komeno T.
Published: November 2008

In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR’s small power output of 300 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR’s simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The internal upper entry CRDs shorten the height of the reactor vessel (RPV) and consequently shorten the primary containment vessel (PCV). The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The Compact Containment Boiling Water Reactor (CCR) has possibilities of attaining both economical and safe small reactor by simplified system and compact PCV technologies.
COBRA Code Assessment for Dry Out Power of Advanced CANDU Fuels
Authors: Daverio Hernando , Juanico´ Luis E. , Delmastro Dari´o
Published: November 2008

COBRA code simulation capability for the thermal hydraulic performance of advanced CANDU fuel element (CANFLEX) was studied. Checking the experimental tests carried out at Stern Lab, the dryout power was predicted by COBRA within a 10% band error for a wide range of conditions. Thus, COBRA was used for comparing the performance of the Argentinean design CARA and the AECL/KAERI design CANFLEX, finding comparable performances.
Improvement in Thermal Margin and Extraction of Maximum Power From Reactor Core for a Given Critical Heat Flux Ratio for an Advanced Heavy Water Reactor
Authors: Lele H. G. , Srivastava A. , Gupta S. K.
Published: November 2008

Critical heat flux is one the important design parameters for nuclear reactor especially for boiling water reactors. It limits power and various operating conditions of a reactor like pressure, flow etc. It is more important for natural circulation boiling water reactor due to its various characteristics. It is necessary to keep margin for CHF in term of Minimum Critical Heat Flux Ratio (MCHFR). Optimisation of such margin improves economic of a reactor. Correlations of critical heat flux are continually being revised as a result of additional experimental data, changes in fuel assembly design, and improved calculation techniques involving coolant mixing and the effect of axial power distributions. There are large variations in different correlation for evaluation of critical heat flux and large numbers of uncertainities are involved. This results in further reduction in margins available. In this paper critical heat flux in a coolant channel of Advanced Heavy Water Reactor is estimated using different correlation at various axial locations. Optimisation study is done by minimizing Minimum Critical Heat Flux Ratio (MCHFR) by use of different axial power profiles. Various parametric studies are done for different number of nodes, different power ratings, thermal hydraulic conditions, different correlations etc.