Showing papers in "Journal of Power and Energy Systems in 2008"

Measurement of Key Pool Boiling Parameters in Nanofluids for Nuclear Applications

Abstract: Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e., contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout.

Two-Phase Swirling Flow in a Gas-Liquid Separator

Abstract: Air-water swirling flows in a one-fifth model of a steam separator in a boiling water nuclear reactor are measured to obtain a database for modeling and verification of numerical methods for predicting swirling flows in the separator. Flow patterns, liquid film thicknesses, separated flow rates and the ratio Ws* of the separated flow to the total liquid flow are measured using a high-speed camera, a laser focus displacement meter and flowmeters. Main conclusions obtained are as follows: (1) liquid transfer from droplets to liquid film is caused not only by droplet deposition but also by the collection of droplets on the vanes of the swirler, (2) Ws* increases with the gas volume flux JG and does not depend on the liquid volume flux JL so much because a large centrifugal force caused by the swirler makes most of droplets in the gas core deposit on the liquid film before the separation and (3) a local peak appears in the axial distribution of film thickness, the position of which corresponds to the location where the droplet deposition caused by the centrifugal force has completed.

Optimal Cycle Scheme of Direct Cycle Supercritical CO2 Gas Turbine for Nuclear Power Generation Systems

Abstract: A supercritical CO2 turbine cycle can achieve a considerably high cycle thermal efficiency at medium turbine inlet temperatures of 500–650°C at high pressure such as 20 MPa, which is too high to produce a reactor pressure vessel within the existing fabrication limits. To solve this problem, a dual expansion turbine cycle is effective; its application was examined for both the fast reactor (FR) of 527°C and 12.5 MPa and a high-temperature gas-cooled reactor (HTGR) of 650°C and 8 MPa. Results showed that, in the case of FR, the cycle thermal efficiency became 42.6%, 44.0%, and 45.1%, respectively, for the 12.5 MPa cycle, the dual expansion cycle, and the 20 MPa cycle. Therefore, the dual expansion cycle is effective. On the other hand, for HTGR, the cycle thermal efficiency became 47.5%, 48.5%, and 50.3%, respectively, for the 8 MPa cycle, the dual expansion cycle, and 20 MPa cycle. In this case, the cycle efficiency advantage becomes smaller than that for the FR, but a 1.0% advantage is obtainable.

Calculation Method for the Prediction of the Performance of a Traveling-Wave Thermoacoustic Cooler

Abstract: When a traveling acoustic wave propagates through a regenerator, the gas in the regenerator undergoes the Stirling thermodynamic cycle, and thus, the energy conversion between heat flux and acoustic power takes place. A cooler that utilizes this energy conversion is called as a traveling-wave thermoacoustic cooler. Swift et al. [The Journal of the Acoustical Society of America, 105, 711 (1998)] have proposed a new traveling wave thermoacoustic cooler that is equipped with a looped tube. This paper describes a numerical method to estimate the performance of this thermoacoustic cooler and shows a comparison between the estimated and experimentally obtained performances.

Creep Rupture Properties of Welded Joints of Heat Resistant Steels

Abstract: In this study, mechanical and creep rupture properties on similar welded joints of Grade 91 steel (Mod.9Cr-1Mo) and dissimilar welded joints of Grade 91 / Inconel82 / SUS304 were examined at elevated temperatures. The effects of temperature and stress on the failure location of welded joints were also investigated. Creep rupture tests were conducted at three temperature levels: 823, 873, and 923 K, the stress ranged from 160 to 240 MPa, 80 to 160 MPa, and 40 to 80 MPa respectively. The creep-rupture strength of the specimen with welded joints was lower than that of the test specimen of base metal at all temperature levels. In addition, the differences in creep strength between the welded joint specimen and the base metal specimen to be greater at higher temperature levels. The fracture Type, observed after long-term creep rupture tests on dissimilar welded joints was transformed from a Type V fracture and a Type VII fracture to a Type IV fracture. It was determined that the fracture type of the 100,000 hours ruptures strength at 823K and 873K for dissimilar welded joint were Type IV fracture.

Turbulent Heat Transfer for Heating of Water in a Short Vertical Tube

Abstract: The turbulent heat transfer coefficients for the flow velocities (u=4.0 to 21 m/s), the inlet liquid temperatures (Tin=296.5 to 353.4 K), the inlet pressures (Pin=810 to 1014 kPa) and the increasing heat inputs (Q0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured by an experimental water loop. The Platinum test tubes of test tube inner diameters (d=3, 6 and 9 mm), heated lengths (L=32.7 to 100 mm), ratios of heated length to inner diameter (L/d=5.51 to 33.3) and wall thickness (δ=0.3, 0.4 and 0.5 mm) with surface roughness (Ra=0.40 to 0.78 μm) are used in this work. The turbulent heat transfer data for Platinum test tubes were compared with the values calculated by other workers' correlations for the turbulent heat transfer. The influence of Reynolds number (Re), Prandtl number (Pr), Dynamic viscosity (μ) and L/d on the turbulent heat transfer is investigated into details and, the widely and precisely predictable correlation of the turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data. The correlation can describe the turbulent heat transfer coefficients obtained in this work for the wide range of the temperature difference between heater inner surface temperature and average bulk liquid temperature (ΔTL=5 to 140 K) with d=3, 6 and 9 mm, L=32.7 to 100 mm and u=4.0 to 21 m/s within ±15% difference.

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

Abstract: Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

Development of Design Technology on Thermal-hydraulic Performance in Tight-lattice Rod Bundles: V-Estimation of Void Fraction

Abstract: An estimation of the void fraction in a tight-lattice rod bundle was needed for the R&D of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa conditions. A spacer effect test was also carried out. The following estimations were conducted: (1) a similarity of the advanced analysis codes with the 3D void fraction data, (2) the comparisons of the TRAC-BF1 code and a drift-flux model with the 1D data. Followings were made clear: (a) The void fraction becomes lower at the peripheral and higher at the rod gap part of the lower core and at the center of the subchannel of the upper core, (b) the codes calculates the similar distribution to the data, and (c) the TRAC-BF1 and the drift-flux model tends to overestimate the void fraction at the lower quality region, on the other hand at the higher quality, those methods tend to same characteristics to the data. It was confirmed that several special features were existed in the tight-lattice rod bundle but the codes were applicable.

The Experimental Study of Atmospheric Stirling Engines Using Pin-Fin Arrays' Heat Exchangers

Abstract: This paper reports experimental results on two kinds of atmospheric Stirling engines that were designed and manufactured using a pin-fin array heat exchanger for the heater and cooler (abbreviated to “pin-fin Stirling engine” hereafter). The first one is a large β type pin-fin Stirling engine with a 1.7-liter displacement volume and power piston volume. The heater consists of an aluminum circular disk with a diameter of 270mm and with large-scale pin-fin arrays carved into the surface. The maximum output reached 91W at a temperature difference of 330K, which is 36% of the scheduled value and 68% of the Kolin's cubic power law. The maximum thermal efficiency was estimated 4.2%. The second engine is an α type pin-fin Stirling engine. Glass syringes were used for the piston-cylinder system and the Ross-yoke mechanism was used for the crank mechanism. By changing temperature difference, the characteristic of output torque in the large range was measured with a precision torque detector.

Measurement of Work Generation and Improvement in Performance of a Pulse Tube Engine

Abstract: Apart from double acting type engines, Stirling engines have either 2 pistons in 2 cylinders or 2 pistons in a single cylinder. Typically, the heater, regenerator and cooler are installed between the 2 pistons. The pulse tube engine, on the other hand, consists of a single piston in a single cylinder, a pulse tube, a heater, a regenerator, a cooler and a second cooler. For this paper, a simple prototype engine that uses air at normal atmospheric pressure as the working gas was fabricated. The oscillating velocity of the working gas in the pulse tube was measured using LDV, and the work flow emitting out of the pulse tube was observed. In addition, the effect of inserting heat storage material in the pulse tube on shaft power and indicated power was examined experimentally. A dramatic increase in the shaft power was achieved.

Calibration and Validation of a Proportional Counter for Determining Beta Emitters

Abstract: Calibration of recently installed proportional counter at the Hot Cells Facility of the Jozef Stefan Institute was performed. Instrument was calibrated for determination of total beta activity, Sr-90 and Pb-210. Detection efficiencies for K-40, Sr-90, Y-90, Pb-210 and Bi-210 were determined, allowing for more accurate determination of the particular nuclide as a single K-40 efficiency. In addition, self-absorption curves for different surface densities for the nuclides mentioned were derived. Two empirical equations for faster and more accurate determination of Sr-90 and Pb-210 were derived. These two equations consider differences in surface density and in-growth of Y-90 and Bi-210, respectively. The detection efficiencies obtained ranged from 10 to 52%, depending on the nuclide, surface density and chemical compositions of the salts used or precipitates obtained following radiochemical separation in the experiment. As a performance test of derived empirical equation for the determination of detection efficiency for Pb-210, specific activity of Pb-210 in IAEA 385 and IAEA 414 intercomparison materials were determined. All procedures and formulae developed include calculation of minimal detectable activities and uncertainty budgets for the determinations concerned.

Development of Small-Scale CHP Plant with a Wood Powder-Fueled Stirling Engine

Abstract: Small-scale biomass CHP (combined heat and power) plants are in demand for environmental reasons - particularly systems fueled by wood waste, which are simple to operate and require no maintenance while having high thermal efficiency similar to oil-fired units. A 55kWe Stirling engine CHP system, combined with a simplified biomass combustion process that uses pulverized wood powder has been developed to meet these requirements. Wood powder of less than 500 μm was mainly used in these tests, and a combustion chamber length of 3 m was applied. Under these conditions, the air ratio can be reduced to 1.1 without increasing CO emissions by less than 10 ppm, and with combustion efficiency of 99.9%. Under the same conditions, NOx emissions are estimated to be less than 120 ppm (on the basis of 6% O2). Wood powder was confirmed to have excellent properties as a fuel for Stirling engines. The 55 kWe Stirling engine performance test was carried out to optimize the operating condition of wood powder burners. The status of Stirling engine operation at a full load with 55 kWe was stable, and start-up and shut -down operations were easy to perform. Operational status was evaluated as being excellent, except for an ash fouling problem in the Stirling engine heater tubes. Ash fouling characteristics were considered in the final stage of the demonstration test. This paper summarizes the wood powder combustion test and Stirling engine performance test. Furthermore, the ash fouling data is shown and the mechanism of ash fouling in heater tubes is discussed.

Numerical Simulation of Thermal Striping Phenomena in T-Junction Piping System Using Large Eddy Simulation

Abstract: In Japan Atomic Energy Agency (JAEA), simulation code “MUGTHES (MUlti Geometry simulation code for THErmal-hydraulic and Structure heat conduction analysis in boundary fitted coordinate)” has been developed to evaluate thermal striping phenomena that are caused by turbulence mixing of fluids in different temperature. MUGTHES employs Boundary Fitted Coordinate (BFC) system to treat complex geometries in power plants. And MUGTHES can deal with three-dimensional transient thermal-hydraulic problem coupled with three-dimensional transient heat conduction in the surrounding structure in consideration of conjugated heat transfer. In this paper, numerical schemes for thermal-hydraulic simulation employed in MUGTHES are described including LES model. A simple method to limit numerical oscillation is adopted in energy equation solving process. A new iterative method to solve Poisson equation in BFC system is developed for effective transient calculations. This method is based on BiCGSTAB method and SOR technique. As the code validation of MUGTHES, a numerical simulation in a T-junction piping system with LES approach was conducted. Numerical results related to velocity and fluid temperature distributions were compared with an existing water experimental data and the applicability of numerical schemes with LES model in MUGTHES to the thermal striping phenomenon was confirmed.Copyright © 2008 by ASME

CFX Simulations of ROCOM Slug Mixing Experiments

Abstract: The present paper documents the CFD code validation activity carried out at the University of Pisa. In particular, the ANSYS CFX-10.0 code was used to simulate one of the experiments conducted at the ROCOM mixing test facility (FZD, Germany), that reproduced the injection of a de-borated slug in one cold leg of a pressurized water reactor (simulated by a salt tracer) with all circulation pumps at steady-state operation. The calculations were run on several grids obtained through different meshing strategies and having different sizes. The numerical results, in terms of normalized concentration of the transported passive scalar in the downcomer and at the core inlet, were compared against corresponding values obtained through experimental measurements of electrical conductivity in the ROCOM facility. Such comparison resulted in a general good qualitative agreement between simulations and experiments, while some discrepancies were evidenced from a quantitative point of view.

SCW Pressure-Channel Nuclear Reactor Some Design Features

Abstract: Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

Abstract: In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters and seismic tests subjected to strong earthquake motions. Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001. This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are developed and calculated. Main criteria for estimating the maximum leakage rate for the lid metallic seal system are no loss of the pre-stress of the lid bolts, no appearance of the plastic region between the metal seal flanges, and no large relative deformation of the lid seals. Finally, in both cases, the low leakage rate for the metal cask lid closure system under the impulsive loads due to aircraft engine crash will be proved thoroughly.Copyright © 2008 by ASME

Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

Abstract: Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Corrosion Resistance of Titanium Alloy on the Overpack for High-level Radioactive Waste Disposal

Abstract: Crevice corrosion of titanium and its alloys were investigated in 10% sodium chloride at 100 °C simulating the environment of the overpack near the seaside. The pH and Chloride ion concentration inside the crevice were monitored by using W/WO3 and Ag/AgCl microelectrode, respectively. The pH and Cl- concentration within the crevice were calculated from the standard potential-pH and potential-log [Cl-] calibration curves. The effect of Mo on the crevice corrosion of titanium was mainly studied. The passivation behavior of the titanium and Ti-15% Mo alloy were also studied using electrochemical impedance studies. A marginal decrease in pH and increase in Cl- ion concentration were observed for pure titanium at 100 °C, where there was large increase of the crevice current. On other hand, there was no apparent change in pH and Cl- ion activity inside the crevice for Ti-15% Mo alloy, where there was no increase of the crevice current. Based on the results, it has been documented that the Ti-15%Mo alloy was not susceptible to crevice corrosion in 10% NaCl solutions at 100 °C. The corrosion reaction resistance (Rt) was found to increase with addition of Mo as an alloying element and also increase with applied anodic potential. Hence, Mo is able to be an effective alloying element, which enhanced the crevice corrosion resistance of titanium under the environment simulating the overpack near the seaside.

Development History and Operation Experience of Ultra-supercritical (USC) Power Plants

Abstract: While significant efforts to prevent further Climate Change have been put into practice, coal is, and coal will always be, playing the major role in global energy supply in next decades. To use coal with minimal environmental impact, one of the practical solutions that are available today is the Ultra-Supercritical (USC) technology which is a well-established technology with high potential for further development in the future. Started in the USA, development of new materials that were applicable to high steam conditions took place predominantly in Europe and Japan. This led to construction of many USC power plants in recent years, and Electric Power Development Co. (EPDC) in Japan has been one of the most experienced USC operators in the world. Further development of the technology is being carried out in Europe, the USA and Japan.

Radioactive Carbon Isotope Monitoring System Based on Cavity Ring-down Laser Spectroscopy for Decommissioning Process of Nuclear Facilities

Abstract: In decommissioning process of nuclear facilities, large amount of radioactive isotopes are discharged as waste. Radioactive carbon isotope (14C) is one of the key nuclides to determine the upper limit of concentration in the waste disposal. In particular, 14C on the graphite reactor decommissioning should be separated from stable carbon isotopes (12C and 13C) and monitored for the public health and safety. We propose an isotope analysis system based on cavity ring-down laser spectroscopy (CRDS) to monitor the carbon isotopes (12C, 13C and 14C) in the isotope separation process for the graphite reactor decommissioning. This system is compact and suitable for a continuous monitoring, because the concentration of molecules including the carbon isotope is derived from its photo absorbance with ultra high sensitive laser absorption spectroscopy. Here are presented the necessary conditions of CRDS system for 14C isotope analysis through the preliminary experimental results of 13C isotope analysis with a prototype system.

Corrosion Behavior of Carbon Steel for the Overpack in the Groundwater Containing Bicarbonate Ions

Abstract: Carbon steel is considered in Japan the candidate material for overpacks in high-level radioactive waste disposal. Effects of bicarbonate solutions on the corrosion behavior and corrosion products of carbon steel were investigated by electrochemical measurements, FT-IR and XRD analyses. The anodic polarization measurements showed that bicarbonate ions (HCO3-) accelerated the anodic dissolution and the outer layer film formation of carbon steel in the case of high concentrations, on the other hand, it inhibited these processes in the case of low concentrations. The FT-IR and XRD analyses of the anodized film showed that siderite (FeCO3) was formed in 0.5 to 1.0mol/L bicarbonate solution, and Fe2(OH)2CO3 in 0.1 to 0.2mol/L bicarbonate solution, while Fe6(OH)12CO3 was formed in 0.02 to 0.05mol/L bicarbonate solutions. The stability of these corrosion products was able to be explained by using the actual potential-pH diagrams for the Fe-H2O-CO2 system.

Effect of Correlations of Component Failures and Cross-Connections of EDGs on Seismically Induced Core Damages of a Multi-Unit Site

Abstract: Aiming at proposing effective applications of seismic probabilistic safety assessment (PSA) for design and risk management of nuclear facilities, we conducted a preliminary seismic PSA study for a multi-unit site to examine core damage frequency (CDF) and core damage sequences with consideration of the effect of correlations of component failures. In addition, we also examined the effectiveness of an accident management measure, namely, cross-connections of emergency diesel generators (EDGs) between adjacent units in this study. Twin BWR-5 units of the same design were hypothesized to be located at the same site in this study and the CDF as well as the accident sequences of this two-unit site were analyzed by using SECOM2, a system reliability analysis code for seismic PSA. The results showed that the calculated CDF was dependent on the assumptions on the correlations of component failures. When the rules for assigning correlation coefficients of component responses defined in the NUREG-1150 program were adopted, the CDF of a single unit, the CDF of this two-unit site (the frequency of core damages of at least one unit at this site) and the frequency of simultaneous core damages of both units increased by factors of about 1.3, 1.2 and 2.3, respectively. In addition, it might be possible that the simultaneous core damages of both units are caused by different accident sequence pairs as well as the same sequence pairs. When cross-connections of EDGs between two units were available, the CDF of a single unit, the CDF of this two-unit site as well as the frequency of simultaneous core damages of both units decreased. In addition, the CDF of this two-unit site was smaller than the CDF of a single unit site. These results show that cross-connections of EDGs might be beneficial for a multi-unit site if the rules for assigning correlation coefficients defined in NUREG-1150 program are reasonable.

Irradiation Test of Fuel Containing Minor Actinides in the Experimental Fast Reactor Joyo

Abstract: The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP accounting for both prompt and delayed heating components, and then adjusted using E/C for B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO2-x or AmO2-x in the (U,Pu)O2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel.

Visualization and Measurement of Dynamic Water Behavior in Polymer Electrolyte Fuel Cell by Neutron Radiography

Abstract: Visualization of water behavior in a polymer electrolyte fuel cell was conducted by neutron radiography. Real-time video images and high-resolution still images were photographed with measurement of fuel cell generation performance. It was clearly visualized by real time imaging that water condensed in an air supply line was coming into a fuel cell. In this case the cell voltage decreased when water flowed into the cathode side. On the other hand, no effect of condensed water into the anode side was observed on the cell performance. In the results of high intensity still imaging, water amount in gas diffusion layer (GDL) under ribs and channel part with GDL were measured quantitatively. The oxygen utilization was varied while PEFC was in a constant current operation. Water amount in the GDL under ribs increased, and cell voltage decreased with increasing oxygen utilization. It can be said that water amount in the PEFC might strongly depend on airflow rate, and the cell voltage might be affected by the water amount.

Heat, Mass and Aerosol Transfers in Spray Conditions for Containment Application

Abstract: TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surete Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Application of Proton Conductors to Hydrogen Monitoring for Liquid Metal and Molten Salt Systems

Abstract: The chemical control of impurity such as hydrogen and oxygen in coolants is one of the critical issues for the development of liquid metal cooled fast reactors and self-cooled liquid breeder blankets for fusion reactors. Especially, hydrogen (isotopes) level is the key parameter for corrosion and mechanical properties of the in-reactor components. For fission reactors, the monitor of hydrogen level in the melt is important for safety operation. The control of tritium is essential for the tritium breeding performance of the fusion reactors. Therefore, on-line hydrogen sensing is a key technology for these systems. In the present study, conceptual design for the on-line hydrogen sensor to be used in liquid sodium (Na), lead (Pb), lead-bismuth (Pb-Bi), lithium (Li), lead-lithium (Pb-17Li) and molten salt LiF-BeF2 (Flibe) was performed. The cell of hydrogen sensor is made of a solid electrolyte. The solid electrolyte proposed in this study is the CaZrO3-based ceramics, which is well-known as proton conducting ceramics. In this concept, the cell is immersed into the melt which is containing the hydrogen at the activity of PH1 of ambient atmosphere. Then, the cell is filled with Ar-H2 mixture gas at regulated hydrogen activity of PH2. The electromotive force (EMF) is obtained by the proton conduction in the electro chemical system expressed as Pt, Melt(PH1) | Proton conductor | PH2, Pt. The Nernst equation is used for the evaluation of the hydrogen activity from the obtained EMF. The evaluations of expected performance of the sensor in liquid Na, Pb, Pb-Bi, Pb-17Li, Li and Flibe were carried out by means of the measurement test in gas atmosphere at hydrogen activities equivalent to those for the melts in the reactor conditions. In the test, the hydrogen activity in the gas varied from 2.2x10-14 to 1. The sensor exhibited good response, stability and reproducibility.

VITRAN: an Advance Statistic Tool to Evaluate Fretting-wear Damage

Abstract: A non-linear vibration model (called VITRAN for VIbration TRansient Analysis-Nonlinear) of the dynamic response of a nuclear fuel rod and its supports has been developed and integrated to a fretting-wear analysis method to predict the performance of fuel assemblies. The approach includes the hydraulic, structural and tribological effects considered to be of sufficient importance. A general description of the software and an example of application are provided in this paper.

Conceptual Reactor Design Study of Very High Temperature Reactor (VHTR) with Prismatic-Type Core

Abstract: The preliminary conceptual design study of prismatic-type Very High Temperature Reactor (VHTR) has been performed with 950°C outlet coolant temperature for higher efficient hydrogen and electricity production. First, the core internals that enable higher outlet temperature are considered in the viewpoint of reduction of core bypass flow. Three-dimensional thermal and hydraulic analyses are carried out and show that the 950°C outlet temperature requires approximately 90% fuel flow fraction and it can be achieved with the installation of the seals in bottom blocks, the coolant tubes in the permanent side reflectors and the core restraint devices. Next, the core and fission product (FP) release analyses are performed. The analysis methods that have been developed for the pin-in-block fuel, one type of prismatic VHTR cores, can be applied to multi-hole fuel, another type of the cores, with some adjustments of the analytical models.