bluuebunny/crossref_metadata_2025 huggingface dataset:
Evaluation of Thermal Comfort and Energy Demands in University Classrooms
Authors: Farooq Sobia , Brown Fredericka
Published: February 2010

The impetus of this study was to evaluate the current HVAC related energy demands of select classroom at The University of Texas at Tyler at present thermal set points and compare the current energy demands with energy demands based on operating the system at the preferred temperature range of occupants. To determine the preferred temperature range of the students at The University of Texas at Tyler, a subjective assessment was performed by questionnaire survey in a selected classroom along with objective measurements of thermal comfort parameters (air velocity, operative temperature and relative humidity). The questionnaire survey included questions about thermal sensation, perception, acceptability, and relevant demographic and clothing data. Using the Fanger’s thermal comfort model, the Predicted Mean Vote (PMV) and Percentage People Dissatisfied (PPD) was calculated from the objective measurements. Regression analysis performed on the survey data provided the range of neutral, preferred and acceptable temperatures in the classroom. The key contributions of this study were: 1) successful implementation of the on field methodology to access thermal comfort in the hot and humid climate of Tyler, Texas, 2) evaluation of the thermal comfort level of the students and faculty at The University of Texas at Tyler, 3) data acquisition of neutral and preferred temperature ranges which can be used as a reference for HVAC design engineers, and 4) comparison of the relationship between thermal comfort level and energy consumption.
A Systematic Investigation of the Effects of Microchannel Width, Depth, and Aspect Ratio on Convective Boiling Heat Transfer and Flow Regimes in Parallel Microchannels
Authors: Harirchian Tannaz , Garimella Suresh V.
Published: February 2010

Experiments are conducted with a perfluorinated dielectric fluid, Fluorinert FC-77, to investigate the effects of channel width, depth, and aspect ratio on flow boiling heat transfer and flow patterns in microchannels. Experiments are performed for a fixed mass flux of 630 kg/m2s with eleven different silicon test pieces containing parallel microchannels of widths ranging from 100 μm to 5850 μm and depths ranging from 100 μm to 400 μm. Flow visualizations are performed using a high-speed digital video camera to determine the flow regimes, with simultaneous local measurements of the heat transfer coefficient and pressure drop. In recent work by the authors [1], it was shown that for a fixed channel depth, the heat transfer coefficient was independent of channel width for microchannels of width 400 μm and larger, with the flow regimes in these microchannels being similar; nucleate boiling was also found to be dominant over a wide range of heat fluxes. The aim of the present study is to expand the range of the microchannel dimensions considered, and specifically to investigate as independent parameters the effects of channel width and depth as well as the aspect ratio and cross-sectional area on boiling heat transfer in microchannels. Flow visualizations and heat transfer results show that the channel cross-sectional area is the important governing parameter determining boiling mechanisms and heat transfer in microchannels.
Numerical Study of Thermophoretic Deposition in Laminar and Turbulent Variable-Property Flat Plate Channel Flow
Authors: Goodwin M. A. , Shinn A. F. , Vanka S. P.
Published: February 2010

Thermophoresis is a significant deposition mechanism for the fouling of exhaust gas recirculation (EGR) coolers. The present paper is motivated by the need to understand and predict the rates of deposition of soot particles in heat exchanger fins used in EGR coolers. This preliminary study considers flat plate channels in the laminar and transitional turbulent regimes. Because of the nano-meter range of the particles, an Eulerian approach is used to determine the particle concentration. Full consideration is given to property variations due to changes in temperature. The inlet gas is at a fixed temperature and the effects of wall temperature are studied for several Reynolds numbers in laminar and transitional turbulent regimes. A finite difference method is used to march downstream in both laminar and transitional turbulent flows. The Launder-Sharma two-equation low-Reynolds-number k-ε model is employed in the numerical simulations to obtain data about flow friction, heat transfer and mass transfer.
Heat Transfer in Adjacent Interacting Impinging Synthetic Jets
Authors: Persoons Tim , O’Donovan Tadhg S. , Murray Darina B.
Published: February 2010

An impinging synthetic jet can attain heat transfer rates comparable to a continuous jet, without net mass input. However it needs a forced cross-flow to supply fresh cooling medium. The vectoring effect of adjacent synthetic jets allows directing the flow by changing the phase between the jets. This study uses the vectoring effect of two adjacent synthetic jets to draw in fresh air, while maintaining high impingement cooling performance. The experimental approach applies infrared thermography and particle image velocimetry to quantify the local convective heat transfer and flow field, respectively. The heat transfer profiles for various phase differences have been compared to the mean flow field and wall-normal velocity fluctuation intensity. For a fixed operating point (stroke length and Reynolds number) and geometry, the cooling performance has been optimised for phase and jet-to-surface spacing, resulting in about 90% enhancement of the maximum and overall cooling rate compared to a single jet, without the need for external cross-flow forcing.
Flow Characteristics of Aluminium Oxide Nanofluids
Authors: Egan Vanessa M. , Walsh Patrick A. , Walsh Edmond J.
Published: February 2010

The current study is an investigation in the flow characteristics of aluminium oxide nanofluids. Relative viscosity measurements are obtained for varying volume fractions using both a rotational viscometer and an experimental setup designed for pressure drop measurements in tubes. The effect of nanoparticle size and preparation method is also investigated as predispersed nanofluids of nominal particle size 10nm and 50nm are compared with each other and with a fluid mixed from Al2O3 nanopowder. Volume fractions of between 1% and 7% were tested. The first method employed to obtain viscosities is based on the Hagen-Poiseuille equation for laminar pipe flow, where pressure drop measurement and flow rate measurements are used to determine relative viscosities of various nanofluids samples. Viscosity measurements were also obtained for a number of solutions on a rotational viscometer and compared to the latter and existing models available in the literature. Overall, it was found during experimentation that the relationship between pressure and flow rate for the various nanofluids was linear indicating that the fluids were Newtonian in nature. An increase in viscosity was recorded for increasing volume fraction; however this was seen to be negligible for volume fractions lower than 1%. Overall it was also seen that both methods of determining relative viscosity were in good agreement. There was not a clear indication of the effect of nanoparticle size on the relative viscosity however the nanofluids formulated from purchased Al2O3 powder resulted in a considerably lower relative viscosity when compared to both nanofluids purchased pre-dispersed from suppliers.
Classification of Condensing Heat Exchangers Performance Data by Gaussian Mixtures
Authors: Pacheco-Vega Arturo , Avila Gabriela
Published: February 2010

We introduce a methodology to extract the regimes of operation from condensing heat exchanger data. The methodology uses a Gaussian mixture clustering algorithm to determine the number of groups from the data, and a maximum likelihood decision rule to classify the data into these clusters. In order to assess the accuracy of clustering technique, experimental data from the literature visually classified as dry-surface, dropwise condensation, and film condensation, are used in the analysis. Though there is a discrepancy between the clustering classification and the visual one, an independent evaluation using artificial neural networks (ANNs) shows that the clustering methodology is able to both find the different regimes of operation and classify the data corresponding to each regime.
The Nuclear Energy Option: Social, Political, and Economic Constraints on Nuclear Development
Authors: Guyon Edwin F.
Published: February 2010

It is proposed that currently existing constraints on nuclear development tend towards social, political, and economic issues. It is further proposed that such constraints, though not specifically or directly identified, exert greater control over the development of peaceful uses of nuclear energy than existing technical constraints. Through the evaluation of these constraints the development of a proposed underground nuclear park in a remote location is proposed as a viable response to these constraints and demonstrates the potential of fast reactors as a major source of base load electrical power.
CFD Simulation of Forced Flow Within the THAI Model Containment
Authors: Zirkel Armin , Doebbener Guido , Laurien Eckart
Published: February 2010

This paper presents the current state of an ongoing analysis and validation of turbulence models for three dimensional numerical simulations (CFD-simulations) within containments of nuclear reactors. A severe accident flow inside a containment could be caused by a leak in the primary circuit of the reactor. It is characterized by different velocities, mass transport, the anisotropy and time dependency of the turbulence field as well as the transition between laminar and turbulent flows. Another issue is the complex geometry of a containment with different rooms and obstacles. The analysis is using the experimental data of the THAI model containment, using the TH18 experiment as the reference case. In order to investigate the numerical error, different advection schemes were used and a grid-dependency study was carried out within a half model of the geometry. On a sufficiently fine grid transient simulations were performed by using the Shear Stress Transport and the Reynolds Stress turbulence models. The results of the simulations are showing different deviations from the experiment. Along with the results, a guide for future work is discussed.
Development and D3 Application of ABWR Feedwater System Model
Authors: Huang Hui-Wen , Shih Chunkuan , Hung Hung-Chih , Chang Kai-Lan , Chen Shu-Chuan , Pan Yu-Jen , Tseng Mao-Sheng , Chen Ming-Huei , Cheng Tsung-Chieh
Published: February 2010

This work developed an Advanced Boiling Water Reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element / three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including inadvertent closure of all turbine control valves and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an Emergency Core Cooling System (ECCS) automatic diverse back-up. In Lungmen Nuclear Power Plant (NPP), a diverse manual initiation means for the High Pressure Core Flooder (HPCF) loop C is designed as the back-up of digitalized Engineered Safety Features Actuation System (ESFAS). If the Motor-Driven Feedwater Pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common cause failure (CCF).
Optimization of a Two-Fluid Hydrodynamic Model of Churn-Turbulent Flows
Authors: Guillen Donna Post , Shelley Jonathan K. , Antal Steven P. , Tselishcheva Elena A. , Podowski Michael Z. , Lucas Dirk , Beyer Matthias
Published: February 2010

A hydrodynamic model of two-phase, churn-turbulent flows is being developed using the computational multiphase fluid dynamics (CMFD) code, NPHASE-CMFD. The numerical solutions obtained by this model are compared with experimental data obtained at the TOPFLOW facility of the Institute of Safety Research at the Forschungszentrum Dresden-Rossendorf. The TOPFLOW data is a high quality experimental database of upward, co-current air-water flows in a vertical pipe suitable for validation of computational fluid dynamics (CFD) codes. A five-field CMFD model was developed for the continuous liquid phase and four bubble size groups using mechanistic closure models for the ensemble-averaged Navier-Stokes equations. Mechanistic models for the drag and non-drag interfacial forces are implemented to include the governing physics to describe the hydrodynamic forces controlling the gas distribution. The closure models provide the functional form of the interfacial forces, with user defined coefficients to adjust the force magnitude. An optimization strategy was devised for these coefficients using commercial design optimization software. This paper demonstrates an approach to optimizing CMFD model parameters using a design optimization approach. Computed radial void fraction profiles predicted by the NPHASE-CMFD code are compared to experimental data for four bubble size groups.
Reliability of Reactor Control Digital Systems
Authors: Yastrebenetsky Mikhail A. , Siora Alexander A. , Tokarev Volodymyr I.
Published: February 2010

In spite of wide expansion of digital systems for NPP control, information about realistic operating reliability measures of these systems is still lacking. The paper is a continuation of report [1] and contains results of analysis of operating reliability of digital control systems used in Ukrainian nuclear power plants. This paper contains: - reliability measures of digital instrumentation and control systems of first generation (designed in 1979–1983); - reliability measures of digital reactor control systems of second generation (designed after 2000) including reasons of failures; - analysis of the reasons of NPP violations due to digital instrumentation and control systems.
Heat Transfer Experiments of Mini-Tube Bank
Authors: Ebihara Yutaka , Katsuta Atsushi , Koizumi Yasuo , Ohtake Hiroyasu
Published: February 2010

Heat transfer and flow behavior in the mini rod bank were examined. The tubes are simulated with a 1 mm diameter nickel wire. The tube bank was composed of the 5×5 square-lattice array and the 5×5 staggered array. The tube banks were arranged in the flow channel of 30 mm wide or 15 mm wide, 15 mm high and 480 mm long. Water was used as the test fluid. A flow rate was varied in the range of the Reynolds number Re = uD/ν of 1 ∼ 800, where D is the tube diameter. The approaching velocity of fluid in the channel was in the range of 0.0036 m/s ∼ 0.68 m/s. Experiments were performed at atmospheric pressure. The measured heat transfer coefficients of the rows after the second row were lower than those of the first row and the difference between those increased as the Reynolds number was increased. The difference turned to decrease around Reynolds number = 50 in the 15 mm wide test section experiments of the square–lattice array and around Reynolds number = 200 in the 30 mm wide test section experiments of the staggered array. The heat transfer coefficients reached back to the first row value around Re = 400 in the former experiments. It was confirmed through the present results and the previous results that the heat transfer in the rear rows is deteriorated by the flow stagnation in the wake region of the preceding rod and the deterioration is recovered as the Reynolds number is increased since the wake region becomes disturbed.
Combined Characteristics of MHD Resistance and Heat Exchange in a HLMC Flow in a Transverse Magnetic Field
Authors: Savinov S. Yu. , Beznosov A. V. , Novozhilova O. O. , Antonenkov M. A.
Published: February 2010

Results of experimental research of lead-bismuth heat-carrier stream velocity structure in the cross-section magnetic field at a varied content of oxygen admixture and characteristics of oxide electroinsulating covers are presented. Experimental studies were carried out for the following operation parameters: the lead-bismuth eutectics temperature T = 400–420 °C; thermodynamic activity of oxygen in the coolant a = 10−4–100; the eutectics flow rate through the experimental part Q = 1.8–3.0 m3/h, the coolant velocity in the experimental part w = 1.0–1.7 m/s; the magnetic induction value B = 0–0.85 T; the Reynolds number Re = (1,6–2,7)·105; the Peclet number Pe = 320–4600, and the Hartmann number Ha = 0–365.
Effect of N Content on Mechanical Properties and Microstructure of Alloy 690
Authors: Liu Kui , Hao Xianchao , Gao Ming , Li Shuo , Li Yiyi , Wang Bofang
Published: February 2010

The microstructures and mechanical properties of nitrogen bearing Alloy 690 have been systematically investigated. Alloy ingots with different N addition, range from 38 to 330wt.ppm, were melted using vacuum induction melting (VIM) plus electro-slag re-melting (ESR) double processing techniques. The forged and hot rolled different N content bars were solid solution treated between 1010°C and 1080°C, thermally treated at 715°C for different state mechanical property testing and microstructure study. Microstructure analysis indicated that nitrogen addition to Alloy 690 can effectively refine the solution treated austenite grains. This may be associated with titanium nitrides pinning the grain boundaries and hindering the grain growth during solid solution treatment. More nitrides, which are identified as TiN, were found on the grain boundaries and in the inside of austenite grains with increasing N contents of the alloy. The carbide precipitation at 715°C showed significant difference identified by SEM. At the level of 38, 100 and 220wt.ppm N, the chromium carbide Cr23C6 distribution on the grain boundaries appeared to be semi-continuous; when the N content reached 330wt.ppm, only few discrete type of carbides were observed. The tension testing results at room temperature of different N content alloys proved that both the ultimate tensile strength (UTS) and the yield strength (YS) enhanced about 50MPa when N content was raised from 38 to 330wt.ppm in this alloy; while the corresponding elongation (EL) and reduction in area (RA) adversely dropped about 5%. Room temperature hardness rose with increasing N content, well matched tensile strength. High temperature tension testing results at the range of 900∼1250°C showed that a severely hot ductility dip, representing by the values of the reduction in area (RA), existed in 300wt.ppm and 100wt.ppm nitrogen containing alloys at the lower end temperature range of 950∼1100°C. However, such ductility dip could be improved when the N content was at 220wt.ppm, and completely eliminated at 38wt.ppm N content. At the higher end temperature rang of 1150∼1250°C, the ductility of all 4 nitrogen bearing alloys did not show significant difference, even though the hot ductility of minimum 38wt.ppm N samples was preferable. Nitrogen content did not affect high temperature strength; the UTS values nearly had no change at the same testing temperature with different nitrogen bearing alloys. The carbide precipitation difference of the thermally treated alloy, induced by N addition, may affect Alloy 690 corrosion properties, which needs to be studied in future. The mechanical properties variation both at room temperature and high temperatures of different nitrogen bearing alloys in this study will be certainly beneficial to determine the practical processing routes of Alloy 690.
Mechanical Resistance in Liquid Lead-Bismuth of 316L/T91 Welded by Electron Beam
Authors: Serre Ingrid , Vogt Jean-Bernard
Published: February 2010

The paper deals with the mechanical strength in air and in liquid lead bismuth at 300°C and 380°C of a T91/316L weld joint assembled by electron beam process. Prior to this, the individual response of each base material has been investigated using the small punch test. The T91 is tested according to several tempering conditions. 316L and T91 tempered at 750°C did not exhibit any sign of liquid metal embrittlement. However, for the T91 steel tempered at 600°C or 500°C, a ductile to brittle transition is induced by liquid lead bismuth. After tempering at 650°C and 700°C, liquid lead bismuth promotes a decrease in mechanical properties and a reduction of the ductility of materials. The load-displacement curves in air of the weld joint are similar to that of the T91 base material. The plastic deformation is mainly concentrated in the T91 part of the weld joint which promotes cracking in this material. Testing in liquid lead bismuth bath results in a reduction in ductility and the formation of brittle cracks. The T91/weld interface is found to be rather resistant as it cracks late in the test and after a large crack propagated in the T91 steel.
Study on Flow Boiling Heat Transfer and Two-Phase Flow Pressure Drop in Flat Mini-Channel
Authors: Koizumi Yasuo , Ohtake Hiroyasu , Sato Ken
Published: February 2010

Flow and the heat transfer characteristics of boiling two-phase flow of water in flat mini-rectangular-channels were examined. The cross-sections tested were 1.0×10 to 0.2×10 mm and the flow channel length was 250 mm. Single phase flow pressure drop was well expressed by the method for the usual size in the present experimental range. Boiling heat transfer of 0.5 mm high and 10 mm wide cross section was similar to that of the usual size. However, that 0.2 mm high and 10 mm wide cross section was a little different from that of the usual size. An increase in the heat flux after the onset of nucleate boiling on the boiling curve is milder than that of the usual size. Thus, the critical heat flux was lower than that of the usual size. Flow patterns observed in the present experiments were a little different from the Baker flow pattern chart. Consistent agreement was not obtained between the present results of the two-phase flow pressure drop and predictions by the methods for the usual size and also for a mini tube. Subcooled boiling was observed widely in the test section. This made it difficult to determine the local conditions such as quality that was necessary to calculate the Lockhart-Martinelli parameter for the two-phase flow pressure drop prediction.
A Practical Example of Simulation Assisted Engineering and System Integrated Test for ABWR RIP-MFG System
Authors: Ochi Hitoshi , Mizuide Atsutoshi
Published: February 2010

In Advanced Boiling Water Reactor (ABWR) the recirculation flow is controlled by Reactor Internal Pumps (RIP). For the new construction ABWR plant, Hitachi-GE applied a Motor Fluid coupling Generator system as a power supply of RIPs (RIP-MFG system). In RIP-MFG system, one MFG supplies electric power to five RIPs and controls the speed of RIPs simultaneously. MFG is the highly reliable power supply system and can sustain power supply to RIPs against grid perturbation due to its large rotational inertia [1]. Application for RIP-MFG system to ABWR is world’s first case although both MFG in BWR and RIP in ABWR are proven technologies. Since the recirculation flow system is the key system determining plant operational performance, Hitachi-GE applied Simulation Assisted Engineering (SAE) for RIP-MFG system to get higher reliability for designing RIP-MFG system and keep a plant constructional schedule due to avoid “trial and error” in the start-up test. The design process was broken down into 3 design phases and 7 design steps as follows. Basic design phase consists of step 1 through step 3. Step 1, a simulation code (RIP-MFG code) which contains electronic and hydraulics models for RIP and MFG was developed. Step 2, basic simulation was performed with RIP-MFG code and general inputs. It was concluded that RIP-MFG system would be basically applicable to ABWR recirculation flow control. Step 3, the System Requirements Specification for RIP-MFG system and the component design specifications were documented based upon the simulation. Detail design phase consists of step 4 through step 6. Step 4, The RIP-MFG components were produced and the component level tests were performed. Step 5, RIP-MFG system integrated tests were performed. The objective of the integrated tests was to verify the system design and the RIP-MFG code. The RIP-MFG code outputs have good agreement with both steady state and transient test data. It has concluded that RIP-MFG code is verified. Step6, the RIP-MFG controller parameters are now being evaluated with the verified code. Final test phase consists of step 7. Step 7, the performances of ABWR with RIP-MFG system will be evaluated finally in the plant start-up test.
Effect of Heater Configurations on Transient Heat Transfer for Various Gases Flowing Over a Twisted Heater
Authors: Shibahara Makoto , Liu Qiusheng , Fukuda Katsuya
Published: February 2010

Forced convection transient heat transfer coefficients were measured for helium gas and carbon dioxide gas flowing over a twisted heater due to exponentially increasing heat input (Q0exp(t/τ)). The twisted platinum plate with a thickness of 0.1 mm was used as test heater and heated by electric current. The heat generation rate was exponentially increased with a function of Q0exp(t/τ). The gas flow velocities ranged from 1 to 10 m/s, the gas temperatures ranged from 313 to 353 K, and the periods of heat generation rate ranged from 46 ms to 17 s. The surface temperature difference and heat flux increase exponentially as the heat generation rate increases with the exponential function. Transient heat transfer coefficients increase with increasing gas flow velocity. The geometric effect of twisted heater in this study shows an enhancement on the heat transfer coefficient. Empirical correlation for quasi-steady-state heat transfer was obtained based on the experimental data. The data for heat transfer coefficient were compared with those reported in authors’ previous paper.
Support for Component Monitoring and Condition-Based Maintenance of Heat Exchangers by Means of Data Validation
Authors: Gocht Ulrike , Wagenknecht Michael , Haenel Jana , Traichel Anke , Klein Michael
Published: February 2010

A validation tool for component monitoring and support of condition-based maintenance for technological systems has been developed and applied to heat exchanger validation and trend evaluation going beyond classical data reconciliation methods.
Student Training Course Using the Experimental Fast Reactor Joyo and Related Facilities
Authors: Aoyama Takafumi , Ito Chikara , Ohkawachi Yasushi , Maeda Shigetaka , Suzuki Soju , Chatani Keiji , Takeda Toshikazu
Published: February 2010

The student training courses using the experimental fast reactor Joyo and related facilities of the Japan Atomic Energy Agency (JAEA) have been initiated to utilize the nuclear facilities and their engineering staffs for the education purpose. The development of the student training course was also strongly supported by the faculty of nuclear engineering of domestic universities whose curriculum has recently been reduced. The program covers the reactor physics test analysis of Joyo core or experiments using the Joyo full-scope training simulator, neutron dosimetry, trace amount of noble gas measurement and chemical analysis of sodium, and the program has started after check and review by the specialists in university education. It is expected to promote the human resource development for the younger generation in nuclear industry, and to strengthen the relation between JAEA and universities in research area.
Molten Salt Reactors: A New Beginning for an Old Idea
Authors: LeBlanc David
Published: February 2010

Molten Salt Reactors have seen a marked resurgence of interest over the past decade, highlighted by their inclusion as one of six Generation IV reactor types. The most active development period however was between the late 1950s and early 1970s at Oak Ridge National Laboratories (ORNL) and any new re-examination of this concept must bear in mind the far different priorities then in place. High breeding ratios and short doubling times were paramount and this guided the evolution of the Molten Salt Breeder Reactor (MSBR) program. As the inherent advantages of the Molten Salt concept have become apparent to an increasing number of researchers worldwide it is important to not simply look to continue where ORNL left off but to return to basics in order to offer the best design using updated goals and abilities. A prime example being the trend towards removal of graphite moderation from the central core, as evident in recent French work on the Thorium Molten Salt Reactor (TMSR) and Russian efforts towards the Molten Salt Actinide Recycler and Transmuter (MOSART). Another major change to the traditional Single Fluid, Molten Salt Breeder Reactor (MSBR) design and the primary subject of this presentation is a return to the mode of operation that ORNL had in mind for the majority of its MSR program. That being the Two Fluid design in which separate salts are used for fissile 233UF4 and fertile ThF4. Oak Ridge abandoned this promising route due to what was known as the “plumbing problem”. It will be shown that a simple yet crucial modification to core geometry can in fact solve this problem and allow the great advantages of the Two Fluid design that ORNL had sought for many years. It will also be shown that this updated design can be started on Low Enriched Uranium with a simple transition to a pure Th-233U cycle which removes the need for shipping proliferation sensitive material and relieves the constrictions on large scale start up due to limited supplies of Pu or 233U. In addition, another promising route laid out by ORNL was simplified Single Fluid converter reactors that could obtain far superior lifetime uranium utilization than LWR or CANDU without the need for any fuel processing beyond simple chemistry control. Updates and potential improvements to this attractive concept will also be explored.
Experimental Study on Liquid Film Dryout Under Oscillatory Flow Conditions
Authors: Yamagoe Yosuke , Goto Taisuke , Okawa Tomio
Published: February 2010

The use of high power density core is one of the promising ways to improve economic efficiency of advanced boiling water reactors. It is however known that in boiling two-phase flows, an increase in power density commonly reduces the margin to the onset of unanticipated flow instability. Hence, in the development of a boiling water reactor of high power density core, ability to predict the occurrence of boiling transition is considered indispensable even when the coolant flow rate is not in the steady state. In the present work, sinusoidal oscillation was applied to the inlet mass flux and the experimental measurement of the critical heat flux was carried out under flow oscillation conditions. It was shown that the critical heat flux decreases monotonically with increased values of oscillation amplitude and oscillation period. These results are consistent with experimental data reported by previous investigators. A simple theory was then proposed to estimate the critical heat flux in oscillatory flow condition. Considering the application to the advanced boiling water reactors, the triggering mechanism of the critical heat flux condition is supposed to be the liquid film dryout in annular two-phase flow regime of high vapor quality. Under the flow oscillation condition, it is expected that long waves are formed on a liquid film due to the time variation of inlet mass flux. Assuming that the wave evolution within a boiling channel is influential in the occurrence of the local dryout of a liquid film, an available nonlinear wave theory was applied to the estimation of critical heat flux under the flow oscillation condition. It was demonstrated that the critical heat fluxes measured under the oscillatory conditions agree with the proposed theory fairly well.
Flow and Mass Transfer in Bends Under Flow-Accelerated Corrosion Wall Thinning Conditions
Authors: Pietralik John M. , Schefski Chris S.
Published: February 2010

The three groups of parameters that affect flow-accelerated corrosion (FAC) are flow conditions, water chemistry, and materials. Nuclear power plant (NPP) data and laboratory tests confirm that under alkaline water chemistry there is a close relationship between local flow conditions and FAC rates in piping components. The knowledge of local flow effects can be useful for developing targeted inspection plans for piping components, predicting the location of the highest FAC rate for a given piping component, and determining what piping components should be replaced. A similar evaluation applies also to FAC in heat transfer equipment such as heat exchangers and steam generators. The objective of this paper is to examine the role of flow and mass transfer in bends under FAC conditions. Bends experience increased FAC rates compared to straight pipes, and are the most common components in piping systems. When the flow effects are dominant, the FAC rate is proportional to the mass flux of ferrous ions, which, in turn, is proportional to the mass transfer coefficient in the flowing water. The mass transfer coefficient describes the intensity of the transport of corrosion products (ferrous ions) from the oxide-water interface into the bulk water. Therefore, this parameter can be used for predicting the local distribution of the FAC rate. The current paper presents plant and laboratory evidence of the relationship between local mass transfer conditions and the FAC rate in bends. It shows correlations for mass transfer coefficients in bends and reviews the most important flow parameters affecting the mass transfer coefficient. The role of bend geometry and, in particular, the short and long radii, surface roughness, wall shear stress, and local turbulence is discussed. Computational fluid dynamics calculations and plant artefact measurements for short-radius and long radius bends are presented. The effect of the close proximity of two bends on FAC rate is also examined based on CANDU™ NPP inspection data and compared with literature data.
Innovative Design for Control Rooms
Authors: Hessler Christian , Scharf Tobias
Published: February 2010

Screen-based human-machine interfaces, having been standard for many years in most process industries, are now also well acknowledged in the main control room of nuclear power plants. Being a standard feature of digital I&C systems, they offer significant benefits and interesting options for control room design, but also constitute challenges for the designer. Attractive opportunities for the designer and operator are for example the minimization of equipment, compared to design of hardwired panels, the reduction of cabling and cable separation issues, the flexibility and increased possibility for customization to specific utilities and operator needs. However, this flexibility brings also new challenges to the design, for new builds as well as for plant modernization projects, such as ensuring overall situation awareness of the operator, in spite the intrinsically serial character of information presentation, and the need to integrate different, even diverse digital human-machine interfaces of the safety and non-safety I&C systems. The reference concept of AREVA NP for all projects involving control room design relies on an overall I&C architecture, based on the TELEPERM XS for safety I&C systems, and Siemens SPPA T2000. SPPA T2000’s OM690 screen-based monitoring and control system is used as the integrated main operator interface, supporting plant operation in all plant states. Additionally control boards are used to implement a separate safety panel satisfying nuclear safety qualification requirements. These components are used to tailor the design of main control room, remote shut-down stations and local control stations, in accordance with licensing requirements, utility expectations and operating staff needs.
Three-Dimensional Core Behaviour Analysis of the Gas-Cooled Fast Reactor (GFR) During Hypothetical Control Assembly Withdrawal Accidents
Authors: Girardin Gae¨tan , Blanchet David , Mikityuk Konstantin , Rimpault Ge´rald , Chawla Rakesh
Published: February 2010

The large 2400 MWth Gas-cooled Fast Reactor (GFR) is one of the most promising advanced reactor concepts currently being investigated within Generation IV. The current work presents the three-dimensional simulation and analysis of GFR core behavior related, primarily, to control assembly (CA) fast movements and/or to accidental withdrawals/ejections. These two categories of events differ in the speed at which the CA is removed. The transient analysis is focused on two different GFR core designs: (1) the reference design (“2004-Core”) and (2) a variant design (“2007-Core”), the main difference being in the core geometry, i.e., in the height-to-diameter (H/D) ratio (2.35 m for the latter core, compared to 1.55 m for the reference design). The reported work firstly concerns the development and benchmarking of coupled, full-core 3D neutron kinetics (NK) and 1D thermal-hydraulics (TH) models for both GFR designs. The models were developed using the TRACE and PARCS codes, which form part of the FAST code system applied at PSI for the generic analysis of advanced fast-spectrum systems. The neutronics is simulated in the nodal diffusion approximation, in conjunction with spatial kinetics. In the TH model, each of the fuel subassemblies (SAs) is modeled individually, so that one can accurately track changes in the thermal-hydraulic-channels surrounding the control assemblies. Consistent nodalization schemes are used in the TH and NK models, in order to provide a 1:1 mapping. Using the developed coupled models, spatial effects have been studied for a variety of hypothetical CA driven transients. Asymmetric CA ejections (speed ∼ 4 cm/s) and withdrawals (speed ∼ 2 mm/s) were analyzed, and large spatial deformations of the power map (up to +21%, relative to average fission power in the TH channels adjacent to the ejected CA) were observed. The analysis has clearly brought out the importance of 3D spatial effects, in the context of the correlation between CA-worths and the loose neutronic coupling of core regions in the case, for example, of a low-H/D core. A sensitivity study of the spatial effects to different parameters (e.g., number of assemblies being ejected and the CA speed) has been conducted, in order to fully characterize the GFR dynamic response. Furthermore, comparisons between the GFR core variants were made. Lower temperatures (fuel, clad and coolant) and lower spatial effects were obtained for the second core. Based upon the results obtained, no major safety-related problem has been found as regard to the CA pattern developed for the GFR. In order to limit the spatial effects, the CAs have to be moved either independently or in symmetric configurations, e.g., three CAs with 120°C symmetry.
Experimental Study on Gas Entrainment Due to Non-Stationary Vortex in a Sodium Cooled Fast Reactor: Comparison of Onset Conditions Between Sodium and Water
Authors: Kimura Nobuyuki , Ezure Toshiki , Miyakoshi Hiroyuki , Kamide Hideki , Fukuda Takeshi
Published: February 2010

An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. In the reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems in this design is gas entrainment at the free surface in the R/V. In most of past studies, water experiments were performed to investigate the gas entrainment in the reactor. It is necessary to evaluate an influence of fluid physical property on the gas entrainment phenomena. In this study, sodium experiments were carried out to clarify the onset criteria of the gas entrainment due to a free surface vortex. Water experiments using a test section in which geometry is the same as that in the sodium tests were also performed. The gas entrainment in water slightly tended to take place in comparison with that in sodium under low velocity conditions. Overall onset condition map on lateral and downward flow velocities in the sodium and water experiments were in good agreement.
A Discrete Equation Method for Two-Dimensional Calculations of Two-Phase Compressible Mixtures
Authors: Leonov A. A. , Chudanov V. V.
Published: February 2010

At present time there are different numerical schemes to solve the hyperbolic equations system for two-phase mixture. Those schemes, mainly, rely on a second order Godunov-type scheme (Godunov, 1969), with approximate Riemann solver for the resolution of conservation equations, and set of nonconservative equations. In this paper we applied the Discrete Equation Method (DEM) processed in Saurel & Abgrall (2003) for two-phase compressible mixtures calculation using exact Riemann solver (Leonov & Chudanov, 2008). Thanks to a deeper analysis of the model, a class of schemes, those are able to converge to the correct solution even when shock waves interact with volume fraction discontinuities, was proposed. Such analysis provides a more accurate estimate of closure terms and an accurate resolution method for the conservative fluxes as well as non-conservative terms even for situations involving discontinuous solutions. The relaxation parameters are determined also, so the resulting model is free of input parameters. The second order accuracy numerical scheme was obtained using an extension of the conventional MUSCL approach. Such method allows extending for multidimensions using splitting procedures. Some common set of 2D numerical tests, calculated with previous issues of two-phase model, is produced. Here a two-phase shock tube with tangential velocity discontinuity, the advection of a square gas bubble in uniform liquid flow, the shock wave interactions with density discontinuities are presented.
Iron-Copper-Nickel Many-Body Potential Consistent With Thermodynamics
Authors: Bonny Giovanni , Pasianot Roberto C. , Castin Nicolas , Terentyev Dmitry , Malerba Lorenzo
Published: February 2010

The Fe-Cu-Ni ternary alloy is of interest for nuclear applications because Cu and Ni are considered to have major effects on the embrittlement under irradiation of reactor pressure vessel steels. To improve our understanding on this phenomenon, large scale atomistic simulations in this model alloy are desirable. For this purpose we develop a ternary Fe-Cu-Ni many-body potential consistent with thermodynamics is developed for the first time. The potential was validated using molecular static and atomistic kinetic Monte Carlo simulations and a qualitative agreement with experiments was established. In particular, Cu precipitates were found to be enriched by Ni on the precipitate surface. Also, the effects diluting the Fe-Cu alloy by Ni on mean precipitate size and density showed similar trends as observed in experiments; i.e. no effect of Ni on the mean precipitate size and an increase in the maximum precipitate density due to the addition of Ni. In absolute terms, agreement with experiment is poor due to the limited box size used in the simulations, as correspondingly discussed.
Characteristics of Direct ECC Bypass Phenomena on the Accident Simulation of Late-Phase Reflood in the APR1400
Authors: Kang Kyoung-Ho , Park Hyun-Sik , Choi Ki-Yong , Cho Seok , Choi Nam-Hyun , Yun Byong-Jo , Baek Won-Pil , Kim Yeon-Sik
Published: February 2010

Because the APR1400 adopts a DVI system, of which the injection nozzle are located 2.1 m above the centerline of the cold leg, ECC bypass is one of the most important technical issues related to the downcomer thermal-hydraulics during the LBLOCA sequences. In this paper, among the ATLAS reflood tests, LB-CL-05, LB-CL-09, and LB-CL-15, which represent the Phase-1, Phase-2, and separate effect test, respectively, were discussed in detail from a ECC bypass characteristics point of view. The LB-CL-05 test is one of parametric effect tests for downcomer boiling in the late reflood phase of a LBLOCA and the LB-CL-09 test is an integral test for the reflood phase of a LBLOCA for investigating the thermal-hydraulic characteristics during an entire reflood period. The LB-CL-15 test was performed as a separate effect test peculiar to the condition of the APR1400. Considering the steam flow rate from the intact cold legs and the system parameters such as the pressure and temperature, the critical void height is calculated in each the tests. The calculated critical void height is significantly lower than the distance from the centerline of the intact cold legs to the top of the water in the downcomer, which confirms that the sweep-out phenomena does not occur in the present tests. ECC bypass has been identified as playing an important role in a depletion of the coolant inventory in the reflood phase of a LBLOCA. Although there were some differences, the fractions of the ECC bypass could be assumed to be about 0.5 during the initial period in the LB-CL-05, 09, and 15 tests.
Blind and Post Test Calculation of the Hydrogen Stratification Experiment HM-2 With the Containment Code System COCOSYS
Authors: Burkhardt Joerg , Koch Marco K.
Published: February 2010

This paper shows the results that have been archieved with the German Lumped-Parameter (LP) Code COCOSYS during the blind and post test calculations of the experiment HM-2. The test was carried out in the frame of the OECD THAI Project and was used for a code benchmark in which 6 CFD and 5 LP- Code contributions were submitted. The main phenomena, the built up and the erosion of an atmospheric stratification, were predicted quite well in the blind calculations with COCOSYS. The post test calculations focus on the sensibility of modeling different zone sizes for the plume trace. It was pointed out that this is not a sensitive parameter unless unphysical small values are used.
Impact of the External Factors in the Nuclear Field: A Comparison Between Small Medium Reactors vs. Large Reactors
Authors: Mancini Mauro , Locatelli Giorgio , Tammaro Serena
Published: February 2010

Several improvements have been made in the nuclear energy sector during the last decade leading to new design for advanced nuclear power plants. Literature presents several studies about the economics of these new Power Plants, however the economic analysis of these plants usually considers only the classical accounts related to Construction, Operation & Maintenance, Fuel and Decommissioning. Beside these accounts there are many factors, from now on named External Factors (e.g. social acceptability, enhanced safety, emergency planning zone reduction, etc.) able to heavily determine the profitability of the investment. This paper presents the differential impact of these External Factors on nuclear technology with different sizes. According to the classification currently in use in the IAEA, small reactors are those with electric generation power lower than 300 MW, while medium sized reactors are those with electric power between 300 and 700 MW [1]. We define “Small Medium Reactors” (SMR) reactors with an electrical output smaller than 700 MW (usually 335 MWe) and as “Large reactors”, (LR) reactors with an equivalent electric power greater than 700 MW. (usually 1340 MWe) Starting from the international literature point of view, the paper provides a list of external factors distinguished in economically quantifiable or not. Two different approaches have been used for their assessment: a monetary ranking and a strategic one. Then, using a Quality Function Deployment approach, a multi-attribute model is introduced to obtain a weight for every external factor, dividing their impacts into three sustainability dimensions (economic, environmental and social). The results show that the new SMR perform better than LR thanks to the smaller size which allows an enhancement of the safety level (which affects the public opinion) and a greater flexibility in the market.