bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of an Air Cooled G Class Gas Turbine (the M501GAC)
Authors: Ai Toshishige , Koeneke Carlos , Arimura Hisato , Hyakutake Yoshinori
Published: February 2010

Mitsubishi Heavy Industries (MHI) G series gas turbine is the industry pioneer in introducing steam cooling technology for gas turbines. The first M501G unit started commercial operation in 1997 and to date, with 62 G units sold, MHI G fleet is the largest steam cooled fleet in the market. The existing commercial fleet includes 35 commercial units with more than 734,000 accumulated actual operating hours, and over 9,400 starts. Upgraded versions have been introduced in the 60 and 50Hz markets (M501G1 and M701G2 respectively). On a different arena, MHI is engaged since 2004 in a Japanese National Project for the development of 1,700°C (3092°F) class gas turbine. Several enhanced technologies developed through this Japanese National Project, including lower thermal conductivity TBC, are being retrofitted to the existing F and G series gas turbines. Retrofitting some of these technologies to the existing M501G1 together with the application of an F class air cooled combustion system will result in an upgraded air-cooled G class engine with increased power output and enhanced efficiency, while maintaining the same 1500°C (2732°F) Turbine Inlet Temperature (TIT). By using an open air cooling scheme, this upgraded machine represents a better match for highly cyclic applications with G class efficiency, while the highly reliable and durable steam cooled counterpart is still offered for more base-loaded applications. After performing various R&D tests, the verification process of the air cooled 60 Hz G gas turbine has moved to component testing in the in-house verification engine. The final verification test prior to commercial operation is scheduled for 2009. This article describes the design features and verification plan of the upgraded M501G gas turbine.
Narrow-Band k-Distribution Database for Atomic Radiation in Hypersonic Nonequilibrium Flows
Authors: Bansal Ankit , Modest Michael F. , Levin Deborah
Published: February 2010

Full-spectrum k-distribution (FSK) and multi-group FSK approaches make it possible to evaluate radiative fluxes at a fraction of the cost needed for line-by-line calculations. However, the required k-distributions need to be assembled from accurate absorption coefficient data for each flow condition, which is computationally expensive. An accurate and compact narrow-band k-distribution database has been developed for the most important species encountered in hypersonic nonequilibrium flow. The database allows users to calculate desired full-spectrum k-distributions through look-up and interpolation. Strategies for k-distribution data generation are outlined. The accuracy of the database is tested by comparing narrow-band mean absorption coefficients and narrow-band emissivities with those obtained from line-by-line calculations. Application of the database to construct full-spectrum k-distributions accurately and efficiently is discussed, and results from a number of heat transfer calculations and cpu-time studies are presented.
Concept of a High Pressure Boiling Water Reactor, HP-BWR
Authors: Reisch Frigyes , Tinoco Hernan
Published: February 2010

Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits: 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel without numerous control rod penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340°C (15 MPa) steam instead of ∼285°C (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Steam separators inside the reactor vessel, and steam dryers together with additional separators can be installed inside or outside the containment -Simple dry containment.
Risk-Informed Safety Margin Characterization
Authors: Hess Stephen M. , Dinh Nam , Gaertner John P. , Szilard Ronaldo
Published: February 2010

The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.
RELAP5 Evaluation vs. Marviken Full Scale Containment Experiments
Authors: Facciolo Luca , Almberger Tomas
Published: February 2010

A model of the Marviken Power Station has been built with the analysis code RELAP5. The model has been tested against the measurement data from the experiments run in the facility under the period August 1972 to April 1973 when a series of full-scale blowdown test have been performed. The aim of the work is the employment and the evaluation of the mono-dimensional code RELAP5 for the containment response analysis. Two different blowdown experiments have been selected and used as comparison for the validation test. The simulations have been performed using models that include the main hydraulic and thermal features described in the actual facility. The heat structures and the pressurizers, used to simulate the spray cooling system, have been introduced and treated in separated cases for a better understanding of the response of the model with regards to the presence of these components. In this way it has been possible to distinguish the effects caused by the heat structures from the results induced by the pressurizers. In this way three configurations have been used for each blowdown, one which describes the mere hydraulic system, a second one with the addition of detailed heat structures and a third one including the spray system. The temperature and the pressure initial values for the simulations are not in a steady state but they reproduce the conditions as described at the beginning of each experiments. The discharge rate and the specific enthalpy are accurately set according to the experimental data. The model response for the global fluid pressure and temperature behaviour is, in general, in good agreement with the experimental local data. The difference between the calculations and the experiments are mainly attributed to the inefficient heat transfer between the non-condensable phase (air) and the walls of the structure: the energy is therefore retained in the gas without the possibility to be released to the surroundings causing higher temperatures and pressures in the fluid itself in comparison with the experiments. Furthermore, a one-dimensional model is not able to predict a good mixing of the fluid in the wetwell as it happens in the reality thanks to the natural convection. The pressure, the temperature, the condensation rate as well as the energy in the drywell and in the wetwell are discussed in the paper and compared with the experimental results. In this paper, due to space limitations, only the results for the hydraulic and the heat structure configurations from one blowdown experiment will be discussed.
Fatigue Flaw Tolerance Evaluation of Pressurizer Surge Line for Plant License Renewal Applications
Authors: Ng C. K. , Hankinson S. F. , Salac M. , Swamy S. A.
Published: February 2010

Evaluation of metal fatigue is one of the identified Time Limited Aging Analyses (TLAA) described in the License Renewal Applications for nuclear power plants. A related area is the evaluation of Environmentally Assisted Fatigue (EAF) as required by the Nuclear Regulatory Commission (NRC) in the United States. The focus is on fatigue sensitive components in power plants selected based on anticipated loading conditions and service experience. The pressurizer surge line was identified as a representative component for the evaluation of EAF for nuclear power plants of both older and newer vintage. A particular location of interest on the surge line is the surge line hot leg nozzle to pipe weld. Typical resolution of EAF concerns in a License Renewal Application requires the incorporation of large Environmental Fatigue Multipliers (Fen) in determining the cumulative usage factors for austenitic stainless steel components. Consequently, the 60-year projected cumulative usage factor at this weld location of interest has the potential to exceed the ASME Section III Code allowable limit of 1.0. For components that fail to demonstrate that the predicted cumulative usage factor is less than 1.0, ASME Section XI Non-Mandatory Appendix L provides guidance for evaluating the component’s fitness for service. The approach is based on a fatigue flaw tolerance evaluation and the implementation of an inspection strategy to demonstrate that growth of a postulated flaw would remain below the allowable flaw size for the component of interest. NUREG/CR-6934 establishes a technical basis for improvements to the initially published version of Appendix L requirements. These improvements have been incorporated in the 2008 Addenda to ASME Section XI Code. One of the improvements is the use of Equivalent Single Crack (ESC) aspect ratio to account for the effects of multiple fatigue crack initiations and the linking of these cracks as they grow to form a single long crack with large aspect ratio. Since the NRC in the United States has not yet endorsed the use of 2008 Addenda to the ASME Section XI Code which contains the latest version of Appendix L, fatigue flaw tolerance analysis is performed in accordance with the requirements in NUREG/CR-6934 for the current license renewal applications. This paper demonstrates the use of the fatigue flaw tolerance approach in accordance with the requirements in NUREG/CR-6934. Typical thermal stratification loading and thermal transients including the use of ESC aspect ratios are considered in the fatigue flaw tolerance evaluation. The results are used to demonstrate fitness for service for a typical 14-inch Schedule 160 pressurizer surge line when the predicted cumulative usage factor exceeds 1.0 for 60 years of plant operation.
Survey on Methodologies for Thermal-Hydraulics System Codes Uncertainty Assessment
Authors: Pourgol-Mohammad Mohammad , Sepanloo Kamran
Published: February 2010

With revising the ECCS licensing rules in 1989, USNRC has allowed the use of “best estimate” thermal-hydraulics computer codes (such as RELAP5 and TRAC) with the requirement that uncertainty analysis accompany the results. Several methodologies have been developed for quantification of uncertainties of such codes. These methodologies are either input-driven or output-driven. They differ definition for the uncertainty range, qualification and quantification steps, types of uncertainty sources considered, methods of assignment of uncertainty distribution or range to various parameters, approach to propagation of uncertainty, and the way the dynamic characteristics of TH codes are handled. The IMTHUA methodology, developed by the authors, is a hybrid approach where an input-driven “white box” approach is augmented with output correction based on experimental results relevant to code output. This paper offers a comparative assessment of IMTHUA and other uncertainty analysis methodologies for thermal-hydraulics transient calculations. The methods will be compared based on their approaches for treatment of input, propagation, code models and correlations, as well as output. Comprehensiveness, approach to data treatment, and interpretation of results are among the criteria for comparison. Several examples are provided to clarify the differences.
Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code
Authors: Kisselev Arcadii E. , Strizhov Valerii F. , Vasiliev Alexander D. , Nalivayev Vladimir I. , Parshin Nikolay Ya.
Published: February 2010

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.
Thermo-Hydraulic Design of Westinghouse 4-Loop Reactor Core by COBRA-EN Code
Authors: Jafari J. , Kalagar B. , Abdi Aghdam E. , D’Auria F.
Published: February 2010

The Westinghouse 4-Loop PWR is a 3411MWth Nuclear Power Plant (NPP). The reactor core consists of 193 fuel assemblies within the core shroud. Each fuel assembly is arranged in 17×17 arrays and includes 264 fuel rods, 24 control rod guide tubes and one instrument tube. The objective of thermal and hydrodynamic design is to safely remove of the generated heat in the fuel without producing excessive fuel temperatures or steam void formations and without approaching the critical heat flux under steady-state operating conditions. This paper presents reactor core and fuel assembly modeling of the Westinghouse 4-Loop NPP using the thermo hydraulic subchannel analysis COBRA-EN code. The results of this modeling are compared with the VIPRE-01 thermal hydraulic code. The study involves the determination of the departure from nucleate boiling ratio (DNBR) in the hot channel of the reactor core, the temperature profiles, heat flux and pressure drop across the hottest channel of the hot assemblies. The obtained results shows that the good agreements are exist between the COBRA-EN and VIPRE-01 thermal hydraulic codes.
Next Generation Safety Analysis Methods for SFRs—(4) Development of a Computational Framework on Fluid-Solid Mixture Flow Simulations for the COMPASS Code
Authors: Zhang Shuai , Morita Koji , Shirakawa Noriyuki , Yamamoto Yuichi
Published: February 2010

The COMPASS code is designed based on the moving particle semi-implicit (MPS) method to simulate various complex mesoscale phenomena relevant to core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The MPS method, which is a fully Lagrangian method, can be extended for fluid-solid mixture flow simulations in a straightforward approach. In this study, a computational framework for fluid-solid mixture flow simulations was developed for the COMPASS code. In the present framework, the passively moving solid (PMS) model, which is originally proposed to describe the motion of a rigid body in a fluid, used to simulate hydrodynamic interactions between fluid and solids. In addition, mechanical interactions between solids were modeled by the distinct element method (DEM). Since the typical time step size in DEM calculation, which uses an explicit time integration scheme, is much smaller than that in MPS calculation, a multi-time-step algorithm was introduced to couple these two calculations. In order to verify the proposed computational framework for fluid-solid mixture flow simulations, a series of experiments of water-dam break with multiple solid rods was simulated using the COMPASS code. It was found that simulations considering only fluid-solid interactions using the PMS model can not reasonably represent typical behaviors of solid rods observed in the experiments. However, results of simulations taking account of solid-solid interactions using DEM as well as fluid-solid ones were in good agreement with experimental observations. It was demonstrated that the present computational framework enhances the capability of the COMPASS code for mesoscale simulations of fluid-solid mixture flow phenomena relevant to CDAs of SFRs. To improve the computational efficiency for fluid-solid mixture flow simulations, it will be necessary to optimize the time step size used in DEM calculations by adjusting DEM parameters based on additional experiments and numerical tests.
Measurement of Steam Generator or Reactor Vessel Moisture Carryover Using a Non-Radioactive Chemical Tracer
Authors: Fournier Robert , Thibodeau Michael , French Charles T.
Published: February 2010

All nuclear power plants are licensed to operate at a specific thermal power. Verification of this power level is done by measuring the change in the enthalpy of the feedwater times the mass flow of the feedwater through each steam generator for pressurized water reactor or through the reactor vessel for boiling water reactors. When performing this measurement, it is necessary to determine the enthalpy of the steam exiting each steam generator or the reactor vessel. Since the exit enthalpy of most units is saturated steam, the state of the steam must be defined in terms of the enthalpy of dry stream minus the fraction of moisture being carried over with the steam times the heat of vaporization. The quality of the steam can be measured using a throttling calorimeter, but a much more accurate means of measurement is with the use of a chemical tracer salt that is only carried over with the main steam by droplets of water. Thus, the fraction of moisture in the steam can be defined as the ratio of concentration of the tracer in the steam divided by concentration of the tracer in the riser of the generator or reactor vessel. If the moisture content of the steam is not measured, one must assume that it is zero, which is conservative, since it provides an upper limit for the thermal power calculation. However, from an operational standpoint, if moisture carryover is present but is not accounted for, the thermal power of the unit will be below its licensed power level causing a loss in electrical output due both a lower thermal power and a reduction in turbine cycle efficiency. Moreover, if there is significant carryover, there is the potential for additional damage to the turbine blading and housing from the impingement of the water droplets. This paper describes how a chemical tracer can be used to measure the moisture content of the steam, considerations for the injection and sample points, plant alignment recommendations and the support required from a utility, when performing this test. The paper will also discuss the potential for long term degradation, the effects of extended power uprates on carryover and the limitations of chemistry guidelines.
Operating Experience of Fast Breeder Test Reactor
Authors: Suresh Kumar K. V. , Ramanathan V. , Srinivasan G.
Published: February 2010

Sodium cooled fast breeder reactors constitute the second stage of India’s three-stage nuclear energy programme, for effective utilization of the country’s limited reserves of natural uranium and exploitation of its large reserves of thorium. The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, is a sodium cooled, loop type fast reactor. Its main aim is to provide experience in fast reactor operation and large scale sodium handling and to serve as a test bed for irradiation of fast reactor fuels and materials. FBTR was built on the lines of the French Rapsodie-Fortissimo reactor, with modifications to make it a generating plant. FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. The steam water system mainly consists of a once-through steam generator, which produces super heated steam at a pressure of 125 bars and temperature of 480° C, feed water system and condensate system. The steam produced is supplied to a condensing turbine coupled to an alternator. The reactor achieved first criticality in Oct 85 with a small core of 22 fuel subassemblies (SA) having a unique carbide fuel rich in Pu. This fuel (called MK-I) was developed and made in India and has a composition of 70% PuC-30% UC. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. In the light of the excellent performance of the carbide fuel, which has endured a burn-up of 155 GWd/t without any clad failure, the core has been gradually expanded by the addition of mark II (55% PuC-45% UC) and MOX (44% PuO2−56% UO2) fuel SA to compensate for the burn-up reactivity loss, and the reactor power has been progressively raised from 10.6 MWt to a maximum of 17.4 MWt. Over the years, several safety related experiments have been conducted. These include natural convection tests and experiments to validate the Failed Fuel Detection System. The challenges faced include a major fuel handling incident, primary sodium leak and reactivity transients. Two major modifications were carried out — one on the Steam Generator Leak Detection System and the other in the steam-water circuit. These helped in improving the campaign availability from less than 50% to more than 90%. The main component limiting the life of reactor is the grid plate supporting the core. The fast flux at the grid plate was measured using Np foils. The residual life of the grid plate has been estimated as 11 effective full power years. The paper presents operating experience of the reactor, performance of the carbide fuel, safety related experiments done in the reactor, various challenges faced, various modifications carried out to improve system reliability and availability and residual life assessment of the reactor.
FFS Issues for Steam Generator Tube (Top of Tubesheet) Intergranular Attack/Stress Corrosion Cracking
Authors: Durance David , Sedman Ken , Roberts John , Gorman Jeff
Published: February 2010

The degradation of Unit 4 SG tubing by IGA/SCC has recently limited both the operating period and end of life predictions for Unit 4 since restart in late 2003. SG4 has experienced the most significant damage with substantial increases in both initiation and growth rates from 2005 through the spring of 2007. A detailed review of the occurrence of circumferential OD IGA/SCC at the RTZ in the HL TTS region of Bruce 4 steam generator tubes has led a conclusion that it is probable that the IGA/SCC has been the result of attack by partially reduced sulfur species such as tetrathionates and thiosulfates during periods of low temperature exposure. It is believed that attack of this type has most likely occurred during startup evolutions following outages as the result the development of aggressive reduced sulfur species in the TTS region during periods when the boilers were fully drained for maintenance activities. The modification of outage practices to limit secondary side oxygen ingress in the spring of 2007 has arrested the degradation and has had significant affects on the allowable operating interval and end of life predictions for the entire unit.
Two-Energy Modeling of Heat Conduction in a Heterogeneous Spherical Fuel Element
Authors: Pessoa Carlos V. , Su Jian , de Oliveira Claudio L.
Published: February 2010

In this work, heat conduction in a typical spherical fuel element (pebble) of a pebble-bed high temperature reactor was studied. The fuel element is composed by a particulate region with spherical inclusions, the UO2 fuel particles (TRISO), dispersed in a graphite matrix. The two energy equation model was applied to the particulate region, generating two macroscopic temperatures, respectively, of the fuel and of the matrix. Analytical solutions are obtained for steady-state heat conduction. Transient analysis was carried out by using the generalized integral transform technique (GITT), which requires low computational efforts and allows a fast evaluation of the two macroscopic transient temperatures of the particulate region. The solution by GITT leads to a system of ordinary differential equations with the unknown transformed potentials, which is solved numerically to obtain the hybrid solution of the original partial differential equation. Numerical results for several testing cases are presented.
Numerical Analysis of Supercritical Water Heat Transfer in Vertical Tube With Different Obstacles
Authors: Zhang Bo , Shan Jianqiang , Jiang Jing
Published: February 2010

Supercritical Water Reactors (SCWRs) are essentially water reactors operating at pressure and temperature above critical point. The heat transfer coefficient is relative low when the bulk temperature is above the pseudo-critical point due to the properties of vapor-like fluid. To obtain better heat transfer characteristics, increasing the fluctuation using obstacles is the conventional method. Heat transfer characteristic in vertical tube with different obstacles is numerically investigated under supercritical condition. Numerical simulation is carried out with commercial CFD code Fluent 6.1 and adaptive grid. The results show that The RNG k-ε model with enhanced wall treatment can obtain a reliable result; the blockage ratio and the local temperature have large influence on the heat transfer enhancement. The influence region and decay trend of obstacles are also studied and compared with existing correlations.
TREQ: Fuel Rod Performance Code for Demanding PWR Plants
Authors: Pastor I. , Doncel N.
Published: February 2010

The search for improvements in nuclear fuel cycle economics results in increasing demands for fuel operation, including higher discharged burnup, and plant manoeuvrability. To achieve these objectives without a reduction of safety margins, new fuel designs and materials that enable enhanced performance capabilities have been developed like the dopped pellets products or the new cladding materials. Besides, fuel rod performance code are incorporating the experience gained in these irradiation conditions. In these regards, the European Fuel Group (EFG) has developed the TREQ code. TREQ is a fuel rod thermal-mechanical code used for the design and safety evaluation of fuel rods for PWR reactors. TREQ benefits from more than 30 year experience coming from the PAD code. The TREQ development was started by Westinghouse, BNFL and ENUSA, the EFG and it is primarily applied to the thermal-mechanical analysis of rods operating at steady-state or slow transient operations as well as to provide fuel inputs, for other codes used in the accident analyses and it has been validated over a wide range of irradiation conditions. Thus, TREQ Fuel Rod Performance Code meets customers’ needs taking advantage of the performance of new products. The code comprises physical models which are developed on the basis of both theoretical considerations and experimental or in-pile data. The physical models of the code apply to Zircaloy-4 and ZIRLO cladding rods containing uranium and gadolinia doped pellets, fabricated according to ENUSA/Westinghouse specification. TREQ code has been licensed in Belgium up to a rod average burnup over 70 MWd/kgU for application on fuel rod design for PWR reactors. Besides, the code has been used to support several LOCA analyses in Belgium and other European countries. In Spain, TREQ was used in the licensing process of the high rod burnup programs. At present, EFG participates in several Research and Development (R&D) programs. The data at extended rod burnup, higher power and higher fuel densities, from theses programs will be incorporated to update models or reduce uncertainties. This paper presents the principal characteristics of the code, the main mechanical models and the strategy of development for the future.
Numerical Solution on Spherical Vacuum Bubble Collapse Using MPS Method
Authors: Tian Wenxi , Qiu Suizheng , Su Guanghui , Oka Yoshiaki
Published: February 2010

In this study, the vacuum bubble collapse in liquid pool has been simulated using MPS code. The liquid is described using moving particles and the bubble-liquid interface was set to be vacuum pressure boundary without interfacial heat mass transfer. The motion and location of interfacial particles can be competent in configurating the topological shape of vacuum bubble. The time dependent bubble diameter, interfacial velocity and bubble collapse time were obtained under wide parametric range. The comparison with Rayleigh’s prediction showed a good agreement which validates the applicability and accuracy on MPS method in solving present momentum problems. The potential void-induced water hammer pressure pulse was also evaluated which is instructive for cavitaion erosion study. The bubble collapse with noncondensable gas has also been simulated and the rebound phenomenon was successfully captured which is similar with vapor-filled cavitation phenomenon. The present study exhibits some fundamental characteristics of vacuum bubble hydrodynamics and it is also expected to be instructive for further applications of MPS method to in complicated bubble dynamics.
Update on Code Case N-755 Revision Class 3 Polyethylene Piping
Authors: Golliet Matthew G.
Published: February 2010

Nuclear plants have received United States Nuclear Regulatory Commission (NRC) approval to use polyethylene pipe in safety related systems such as essential service water systems. ASME Code Case N-755 is the basis for the utility requests. ASME Nuclear Code committees are developing a revision to provide clarification to the code case requirements and address NRC concerns. Polyethylene pipe replacement projects provide a solution to the age degradation issues such as corrosion and fouling of buried carbon steel pipe.
Weldability of Neutron Irradiated Stainless Steels by YAG Laser and Low-Heat-Input TIG Welding Techniques
Authors: Sato Masatoshi , Nakata Kiyotomo , Ozawa Masayoshi
Published: February 2010

Type 304 and 316L stainless steels (SS) were irradiated at ambient temperature in the U.S. Advanced Test Reactor up to 2×1025n/m2 (E>1MeV). Weldings were performed under various weld heat-input levels from 1 to 10 kJ/cm for specimens containing up to 29 appm helium (He). After the welding, cross-sectional observation and mechanical property examinations such as tensile and side bend tests were carried out. In the case of the welding with a heat-input of 1 kJ/cm, no weld cracks were found in irradiated 304SS with all of He concentrations tested, and in the case of 2 kJ/cm, no weld cracks were also done in up to 9 appmHe 304SS. On the other hand, 316LSS contained only 0.5 appmHe was cracked by welding at 2 kJ/cm. These results clearly suggested that susceptibility of weld cracking in irradiated 316LSS was remarkably higher than that in 304SS. The mechanical properties were in compliance with their acceptance criteria in the weld joints without cracks. The results of SEM and TEM observations showed these weld cracking was attributed to He cavities formed along grain boundaries, which were made from transmuted He atoms aggregated by tensile stress and heating during welding. In the same He concentration and weld conditions, the cavity density in 316L was about one order magnitude higher than that in 304SS. The computer simulation method for irradiated SS welding was developed based on the He aggregation model during welding. The sound repair-welding conditions, which could be applied actual plants, were proposed using the simulation method.
Volume Reduction of Radioactive Concrete Wastes by Thermal and Mechanical Treatments
Authors: Choi Wang Kyu , Min Byung Youn , Lee Kune Woo
Published: February 2010

The decommissioning of nuclear facilities produces a large amount of radioactive wastes and most of them are concrete waste. In Korea, two decommissioning projects such as the decommissioning of the retired research reactors (KRR-1 & 2) and a uranium conversion plant (UCP) at the Korea Atomic Energy Research Institute (KAERI) has been carried out. By dismantling KRR-2, more than 260 tons of radioactive concrete wastes were generated among the total 2,000 tons of concrete wastes and more than 60 tons of concrete wastes contaminated with uranium compounds have been generated for the UCP decommissioning up to now. The volume reduction and recycling of the wastes is essential to reduce the waste management cost with expectation that an approximate disposal cost for the low level radioactive waste will be more than 5,000 US dollars per 200 liter waste drum in Korea. Therefore, we have developed the volume reduction technology applicable to an activated heavy weight concrete generated by dismantling KRR-2 and the uranium contaminated light weight concrete produced from the UCP decommissioning. We have investigated the characteristics of a separation of the aggregates and the distribution of the radioactivity into the aggregates from the volume reduction point of view using an activated heavy weight concrete taken from the KRR-2 and uranium contaminated light weight concrete from the UCP. It could be concluded that the radioactivity is mainly concentrated in the porous fine cement powder below the size of 1mm and the appropriate heating temperature for the treatment of concrete waste is in the range of 450°C to 500°C. The volume reduction rate could be achieved above 70% by a heating followed by a mechanical separation process. A fine cement radioactive produced during the course of the mechanical treatment process was immobilized and reduced in volume by a slagging using a direct current graphite arc melting system which resulted in a volume reduction factor of about 3.4.
Analysis of Thermal Effects in a Cavitating Orifice Using Rayleigh Equation and Experiments
Authors: De Giorgi Maria Grazia , Bello Daniela , Ficarella Antonio
Published: February 2010

The focus of this research is the experimental and analytical study of the cavitation phenomena in internal flows in presence of thermal effects. Experiments have been done on water and nitrogen cavitating flow in orifices. Transient growth process of the cloud cavitation induced by flow through the throat is observed using high-speed video images and analyzed by pressure signals. Cavitation of thermo-sensible fluid, as cryogenic fluid, presents additional complexities (as compared to that in water) because thermal effects are important. The different cavitating behavior at different temperature and different fluid is related to the bubble dynamic inside the flow. To investigate possible explanations for the influence of fluid temperature on cavitating internal flow, initially, a steady, quasi-one dimensional model has been implemented. The nonlinear dynamics of the bubbles has been modeled by Rayleigh-Plesset equation. In the case of nitrogen, thermal effects in the Rayleigh equation are taken into account by considering the vapor pressure at the actual bubble temperature Tc, which is different from the liquid temperature T far from the bubble. A convective approach has been used to estimate the bubble temperature. The quasi-steady one dimensional model can be extensively used to conduct parametric studies useful for fast estimation of the overall performance of any geometric design. For complex geometry, three-dimensional CFD codes are necessary. In the present work comparison have been done with numerical predictions by the CFD Fluent code in which a simplified form of the Rayleigh equation taking into account thermal effects has been implemented by external user routines.
Fatigue Capacity of Stiffener to Web Frame Connections
Authors: Lindemark Torbjo̸rn , Lotsberg Inge , Kang Joong-Kyoo , Kim Kwang-Seok , Oma Narve
Published: February 2010

Daewoo Shipbuilding & Marine Engineering Co., Ltd. (DSME), StatoilHydro and DNV established a common project to investigate the reason for the difference between calculated fatigue lives and the in-service experience and to assess the fatigue capacity of stiffener web connections subjected mainly to web frame shear stresses. The main objective of the work was to establish fatigue test data and perform numerical analysis of collar plate connections in order to provide improved confidence in analysis methodology for fatigue life assessment. Large scale fatigue tests of different types of connections were carried out to obtain fatigue test data of collar plate connections. Finite element analyses were carried out for comparison with fatigue test data and with measured stresses on the test model. Based on this work recommendations on fatigue design analysis of connections between stiffeners and web frames have been derived. The background for this is presented in this paper.
Experimental Investigation on Immersion of Tunnel Element
Authors: Chen Zhijie , Wang Yongxue , Wang Guoyu , Hou Yong
Published: February 2010

The experiment is carried out to investigate the motion responses of the tunnel element in the immersion under wave actions and the tensions acting on the controlling cables of its immersing system. The CCD camera is utilized in the experiment to record the motion displacements of the tunnel element during its interaction with waves. The time series of motion responses, i.e. sway, heave and roll of the tunnel element and cable tensions are presented. The results of motion responses of the tunnel element against relative wave height H/h, relative width B/L and relative immersing depth d/h are given. Furthermore, the measured cable tensions are analyzed and the phase relation between waves and cable tensions is also discussed.
Design of Laterally Loaded Piles With Bulge
Authors: Du¨hrkop Jan , Grabe Ju¨rgen
Published: February 2010

Several investigations on the behaviour of bulged piles have shown the benefit of this construction. Bulge means the application of vertical steel plates somewhere near ground surface to improve the lateral bearing capacity. This paper contains the description of a modified P-y curve method which is capable of calculating piles with bulge. Based on this procedure hundreds of different monopile foundations for offshore wind power plants are analysed and results are used to create design charts. Therefore the Characteristic Load Method (CLM) is expanded to considere the effect of a bulge.
Calculation of Wave Forces Acting on a Cylinder With a Porous Plate Fixed Inside
Authors: Bao Weiguang , Zhao Fenfang , Kinoshita Takeshi
Published: February 2010

To evaluate wave forces and to estimate the motion of breakwater, a circular cylinder is investigated based on the linear wave theory in the present work. The cylinder possesses a porous sidewall, an impermeable bottom and a horizontal porous plate inside that is fixed in the cylinder to work as obstruct and make wave dissipation more effectively. To simplify the problem, the Darcy’s fine-pore model is applied to the boundary condition on the porous body surface. The boundary value problem is solved by means of the eigen-function expansion approach. The fluid domain is divided into three regions and different eigen-function series are used. The so-called dispersion relation for the region inside the cylinder is quite different from a conventional one due to the existence of the porous plate. It leads to eigen values of complex number. To obtain solutions for the radiation problems, particular solution should be constructed to take account of the normal velocity appearing on the porous boundary. The wave loads are evaluated by integrating the pressure difference on two sides of the wetted body surface. The theoretical works are in good consistence with the experimental results. The Haskind relations are examined for the porous body. It is found that the damping coefficient consists of two parts. In addition to the component of conventional wave-radiating damping, exists a second component caused by the porous effects.
Numerical Modeling and Ocean Testing of a Direct-Drive Wave Energy Device Utilizing a Permanent Magnet Linear Generator for Power Take-Off
Authors: Elwood David , Schacher Al , Rhinefrank Ken , Prudell Joe , Yim Solomon , Amon Ean , Brekken Ted , von Jouanne Annette
Published: February 2010

For the past several years an inter-disciplinary research group at Oregon State University (OSU), working in conjunction with Columbia Power Technologies (CPT) has been researching innovative direct-drive wave energy systems. These systems simplify the conversion of wave energy into electricity by eliminating intermediate energy conversion processes. In support of this research OSU and CPT have developed a hybrid numerical/physical modeling approach utilizing a large scale linear test bed (LTB), and a commercial coupled analysis tool. This paper will present an overview of this modeling approach and its application to the design of a 10kW prototype wave energy conversion system that was tested in the open ocean in the fall of 2008. The data gathered during ocean testing was used to calibrate the numerical model of the device and predict the energy capture potential of the system.
An Experimental Study of Wave-Induced Set-Up Over a Horizontal Reef With an Idealized Ridge
Authors: Yao Yu , Lo Edmond Y. M. , Huang Zhenhua , Monismith Stephen G.
Published: February 2010

Wave action has been the primary focus of near-shore hydrodynamic studies over decades. The wave-induced set-up due to wave breaking is one of the important factors to consider in determining both water level and mass transport above the reef-top, which has ecological as well as engineering significance. Previous investigations into reef hydrodynamics (including laboratory experiments, field investigations, theoretical analysis and numerical simulations) focused mainly on wave-induced set-up on reefs with a horizontal reef-top (Gourlay, 1996. J. Coastal Eng. 27:161–193). It has been observed that a ridge (reef crest) may be present at the reef edge, but so far we are not aware of any published studies on the effects of this type of ridge on the wave-induced set-up over the reef-top. To understand the role of the ridge in wave-breaking mechanism and wave-induced set-up over the reef, a series of experiments were carried out in a wave flume of 36m long and 0.55m wide, with idealized reef-ridge models being installed at the reef edge to simulate fringing reefs with rectangle ridges. The surface elevations at four locations over the reef were measured with Ultralab sensors (General Acoustics), revealing the variation of wave-induced set-up along the reef. Experimental results are reported for two water depths and eighteen regular wave conditions. Also discussed are the effects of the ridge width on the wave-induced set-up. The focus of this paper will be given to the comparison of wave-induced set-ups with and without the rectangular ridges. Preliminary analysis shows that the ridge controls the water level above the reef-top in a way similar to that a broad crest weir controls the water level in open channel flows. Furthermore, the presence of the ridge is found to alter significantly the wave transformation process near the reef edge, especially the strength of the reflected waves and the locations of breakers. Experiments also show that the wave-induced set-up over reef-top with a ridge is generally much larger than that without a ridge. Finally, an attempt is made to introduce a new dimensionless parameter in order to take the ridge configuration into consideration and achieve a better agreement between experiments and predictions when ridges are present.