Journal of Power and Energy Systems
Online ISSN : 1881-3062
ISSN-L : 1881-3062
Volume 3 , Issue 1
Special Issue on 16th International Conference on Nuclear Engineering
Showing 1-29 articles out of 29 articles from the selected issue
Special Issue on 16th International Conference on Nuclear Engineering
Papers
  • Yoshinobu NAKAO, Eiichi KODA, Toru TAKAHASHI
    2009 Volume 3 Issue 1 Pages 2-11
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    We have developed the general-purpose software by which static thermal characteristic of the power generation system is analyzed easily. This software has the notable features as follows.
    -It has the new algorithm to solve non-linear simultaneous equations to analyze the static thermal characteristics such as heat and mass balance, efficiencies, etc. of various power generation systems.
    -It has the flexibility for setting calculation conditions.
    -It is able to be executed on the personal computer easily and quickly.
    We ensured that it is able to construct heat and mass balance diagrams of main steam system of nuclear power plant and calculate the power output and efficiencies of the system. Furthermore, we evaluated various heat recovery measures of steam generator blowdown water and found that this software could be a useful operation aid for planning effective changes in support of power stretch.
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  • Jun MANABE, Yasuhiko SHODA, Tatsushi YAMAMURA, Yuuichiro KUSUMOTO
    2009 Volume 3 Issue 1 Pages 12-22
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Kyushu Electric Power Company Genkai #1 and #2 are twin 500 MW class first generation PWR power stations starting their commercial operation in 1975-1981. The units were recently altered their secondary water treatment from AVT to HAVT (High All Volatile Treatment) operation aiming to suppress erosion in piping and equipment, resulting in both feed water iron concentration reduction to around 1 ppb as an indication of the effects and scale adhesion reduction of feed water pumps and feed water heaters.
    The units had been successfully operated from the start of their commercial operation except for scale adhesion to SG and others, degradation of copper alloy material tubes in auxiliary heat exchangers and lower condenser vacuum derived from protective ferrous sulfate coating.
    Life cycle management program was implemented resulting in the alteration of water treatment to HAVT adopting both the SG blow down demineralizing and the replacement of copper alloy tube heat exchangers to stainless steel and titan tubes eliminating the copper materials.
    Further more the scale adhesion mechanism was introduced of the high temperature region of the secondary system based on field examination for iron characterization of both AVT and of HAVT in Genkai units, confirming HAVT advantageous effect for prevention of the scale adhesion.
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  • Toshiyasu NISHIMURA, Junpha DONG
    2009 Volume 3 Issue 1 Pages 23-30
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    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.
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  • Ikuo IOKA, Chiaki KATO, Kiyoshi KIUCHI, Junpei NAKAYAMA
    2009 Volume 3 Issue 1 Pages 31-37
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Austenitic stainless steels suffer intergranular attack in boiling nitric acid with oxidants. The intergranular corrosion is mainly caused by the segregation of impurities at the grain. An extra high purity austenitic stainless steel (EHP alloys) was developed with conducting the new multiple refined melting technique in order to suppress the total harmful impurities less than 100ppm. The corrosion behavior of type 310 EHP alloy with respect to nitric acid solution with highly oxidizing ions (boiling 8kmol/m3 HNO3 solutions containing 1kg/m3 Cr(VI) ions) was investigated. The straining, aging and recrystallizing (SAR) treated type 310 EHP alloy showed superior corrosion resistance for intergranular attack than solution annealed (ST) type 310 EHP alloy with same impurity level. Boron segregation at the grain boundary was detected in only ST specimen using a Fission Track Etching method. It is believed that the segregated boron along the grain boundaries in type 310 EHP alloy was one of main factor of intergranular corrosion. The SAR treatment was effective to restrain the intergranular corrosion for type 310 EHP alloy with B less than 7ppm.
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  • Takeshi OGAWA, Motoki NAKANE, Kiyotaka MASAKI, Shota HASHIMOTO, Yasuo ...
    2009 Volume 3 Issue 1 Pages 38-50
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    This paper describes the effect of large pre-strain on very high cycle fatigue strength of austenitic stainless steels that are widely used in nuclear power plants. Fatigue tests were carried out on strain-hardened specimens. The material served in this study was type SUS316NG. Up to ±20% pre-strain was introduced to the materials, and the materials were mechanically machined into hourglass shaped smooth specimens. Some specimens were pre-strained after machining. Experiments were conducted in ultrasonic and rotating-bending fatigue testing machines.
    The S-N curves obtained in this study show that an increase in the magnitude of the pre-strain increases the fatigue strength of the material and this relationship is independent of the type of the pre-strain of tension or compression. Although all specimens fractured by the surface initiated fatigue cracks, one specimen fractured by an internal origin. However, this internal fracture did not cause a sudden drop in fatigue strength of type SUS316NG. Vickers hardness tests were carried out to ascertain the relationship between fatigue strength and hardness of the pre-strained materials. It was found that the increase in the fatigue limit of the pre-strained materials strongly depended on the hardness derived from an indentation size equal to the scale of stage I fatigue cracks.
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  • Takehisa HINO, Masataka TAMURA, Yoshimi TANAKA, Wataru KOUNO, Yoshinob ...
    2009 Volume 3 Issue 1 Pages 51-59
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Stress corrosion cracking (SCC) has been reported at the aged components in many nuclear power plants. Toshiba has been developing the underwater laser welding. This welding technique can be conducted without draining the water in the reactor vessel. It is beneficial for workers not to exposure the radiation. The welding speed can be attaining twice as fast as that of Gas Tungsten Arc Welding (GTAW). The susceptibility of SCC can also be lower than the Alloy 600 base metal.
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  • Yoshio SUZUKI, Akemi NISHIDA, Fumimasa ARAYA, Noriyuki KUSHIDA, Taku A ...
    2009 Volume 3 Issue 1 Pages 60-71
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Center for computational science and e-systems of Japan Atomic Energy Agency is carrying out R&D in the area of extra large-scale simulation technologies for solving nuclear plant structures in its entirety. Specifically, we focus on establishing a virtual plant vibration simulator on inter-connected supercomputers intended for seismic response analysis of a whole nuclear plant. The simulation of a whole plant is a very difficult task because an extremely large dataset must be processed. To overcome this difficulty, we have proposed and implemented a necessary simulation framework and computing platform. The computing platform enables an extra large-scale whole nuclear plant simulation to be carried out on a grid computing platform called ITBL-IS, Information Technology Based Laboratory Infrastructure and AEGIS, Atomic Energy Grid Infrastructure. The simulation framework based on the computing platform has been applied to a linear elastic analysis of the reactor pressure vessel and cooling systems of the nuclear research facility, HTTR. The simulation framework opens a possibility of new simulation technologies for building a whole virtual nuclear plant in computers for virtual experiments.
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  • Koji SHIRAI, Kosuke NAMBA, Toshiari SAEGUSA
    2009 Volume 3 Issue 1 Pages 72-82
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    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(1) and seismic tests subjected to strong earthquake motions(2). 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(3)-(6). 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.
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  • Toru OUMAYA, Akira NAKAMURA, Daisuke ONOJIMA, Nobuyuki TAKENAKA
    2009 Volume 3 Issue 1 Pages 83-102
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    The pressurizer spray line of PWR plants cools reactor coolant by injecting water into pressurizer. Since the continuous spray flow rate during commercial operation of the plant is considered insufficient to fill the pipe completely, there is a concern that a water surface exists in the pipe and may periodically sway. In order to identify the flow regimes in spray line piping and assess their impact on pipe structure, a flow visualization experiment was conducted. In the experiment, air was used substituted for steam to simulate the gas phase of the pressurizer, and the flow instability causing swaying without condensation was investigated. With a full-scale mock-up made of acrylic, flow under room temperature and atmospheric pressure conditions was visualized, and possible flow regimes were identified based on the results of the experiment. Three representative patterns of swaying of water surface were assumed, and the range of thermal stress fluctuation, when the surface swayed instantaneously, was calculated. With the three patterns of swaying assumed based on the visualization experiment, it was confirmed that the thermal stress amplitude would not exceed the fatigue endurance limit prescribed in the Japanese Design and Construction Code.
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  • Hideo MACHIDA, Manabu ARAKAWA, Norimichi YAMASHITA, Shinobu YOSHIMURA
    2009 Volume 3 Issue 1 Pages 103-113
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Risk-Informed integrity management methodologies have been developed for Japanese nuclear power plants. One of the issues of concern is the reliability assessment of piping with flaws due to stress corrosion cracking (SCC). Therefore, the probabilistic fracture mechanics analysis code has been developed, which can perform the reliability assessment for austenitic stainless steel piping with flaws due to SCC. This paper describes technical basis of this code. This method is based on Monte-Carlo technique considering many sample cases in a piping section, where the initiation and growth of cracks are calculated and piping failures, including leaks and rapture, are evaluated. A notable feature is that multiple cracks can be treated, consequently, assessment of coalescence of cracks and intricate break evaluation of piping section have been included. Moreover, the in-service inspection (ISI) and integrity evaluation by Fitness-for-Service (FFS) code are integrated into the analysis, and the contribution to failure probability decrease can be assessed. Key parameters are determined on a probability basis with the designated probability type throughout the procedure. Size, location and time of crack initiation, coefficients of crack growth due to SCC and factors for piping failure are included in those parameters. With this method the reliability level of the piping through the operation periods can be estimated and the contribution of various parameters including ISI can be quantitatively evaluated.
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  • Hiroo KONDO, Takuji KANEMURA, Hirokazu SUGIURA, Nobuo YAMAOKA, Mizuho ...
    2009 Volume 3 Issue 1 Pages 114-125
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    This paper reports an experimental study of a liquid lithium (Li) target for the International Fusion Materials Irradiation Facility (IFMIF). Experiments on the Li jet were conducted at a Li loop facility. The test section was a 10-mm-thick and 70-mm-wide flat plane jet along a straight channel. The Li jet free surface was photographed using a CCD camera and a stroboscopic light source. To investigate free surface wakes, measurement of the thickness distribution, analyses of the wake, and numerical calculations were conducted and compared. Results demonstrated that the generation of the wake depended strongly on wettability between Li and the structural material, which is 304 SS; the analytical model described the shape well. From the viewpoint of IFMIF design, the surface wake generated at the nozzle corner had a negligible effect on the surface shape of the D+ beam irradiation region in IFMIF.
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  • Noboru KOBAYASHI, Takashi OGAWA, Shigeo OHKI, Tomoyasu MIZUNO, Takanar ...
    2009 Volume 3 Issue 1 Pages 126-135
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650 °C, and the maximum fuel pin bundle pressure drop of 0.4MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40.
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  • Koji FUJIMURA, Akira SASAHIRA, Junichi YAMASHITA, Tetsuo FUKASAWA, Kun ...
    2009 Volume 3 Issue 1 Pages 136-145
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.
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  • Mitsuhiro SUZUKI, Takeshi TAKEDA, Hideo NAKAMURA
    2009 Volume 3 Issue 1 Pages 146-157
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.
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  • Morimasa NAITO, Yuya SAITO, Kenji TANAI, Mikazu YUI
    2009 Volume 3 Issue 1 Pages 158-169
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    An experimental approach is introduced for understanding how the engineered barriers of a deep geological repository system are affected by fault movement. The experiments are conducted using laboratory simulation test equipment. So far, the experiments indicate that the metal overpack is rotated within the bentonite buffer due to its plasticity, but not breached. Numerical modeling is also developed to supplement the range of the experiments, which is limited by the capability of the test equipment.
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  • Takatoshi HIJIKATA, Tadafumi KOYAMA
    2009 Volume 3 Issue 1 Pages 170-181
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Pyrometallurgical reprocessing technology is currently being focused in many countries for closing actinide fuel cycle because of its favorable economic potential and an intrinsic proliferation-resistant feature due to the inherent difficulty of extracting weapons-usable plutonium. The feasibility of pyrometallurgical reprocessing has been demonstrated through many laboratory scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high-temperature fluids. In this study, a salt transport test rig was installed in an argon glove box with the aim of developing technologies for transporting molten salt at approximately 773 K. The gravitation transport of the molten salt at approximately 773 K could be well controlled at a velocity from 0.1 to 1.2 m/s by adjusting the valve. Consequently, the flow in the molten salt can be controlled from laminar flow to turbulent flow. It was demonstrated that; using a centrifugal pump, molten salt at approximately 773 K could be transported at a controlled rate from 2.5 to 8 dm3/min against a 1 m head.
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  • Satoru MOMOKI, Kaoru TOYODA, Takashi YAMADA, Toru SHIGECHI, Tomohiko Y ...
    2009 Volume 3 Issue 1 Pages 182-193
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    A method of predicting the overall heat transfer coefficient and the temperature at the lower limit of film boiling for a finite-length cylinder with flat top and bottom surfaces has been researched and proposed in a previous paper. This paper presents and compares an analysis in the case of a cylinder with a hemispherical bottom. The film boiling heat transfer around a vertical silver cylinder with a convex hemispherical bottom surface is investigated both experimentally and analytically in the present study. The obtained results are also compared and discussed with the authors' previous results for a finite-length cylinder with flat top and bottom surfaces. Quenching experiments were performed using silver cylinders in saturated water. The diameter and length of the test cylinders are 32mm and 48mm, respectively. The test cylinder was heated up to about 600°C in an electric furnace and then cooled down in saturated quiescent water at atmospheric pressure. The resultant cooling and boiling curves and photographs of the film boiling phenomena are presented and discussed. The average heat transfer performance of the hemispherically bottomed cylinder is about 20% higher than that of the flat bottomed cylinder. The degree of wall superheating at the lower limit of film boiling is about 133K. The saturated film boiling heat transfer around the vertical finite-length cylinder with a convex hemispherical bottom was analyzed by taking into account the convective heat transfers from the bottom, side and top surfaces of the cylinder. The resulting analytical data correlated closely with the experimental data in the present study.
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  • Takeharu MISAWA, Hiroyuki YOSHIDA, Hidesada TAMAI, Kazuyuki TAKASE
    2009 Volume 3 Issue 1 Pages 194-203
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    The three-dimensional two-fluid model analysis code ACE-3D is developed in Japan Atomic Energy Agency for the thermal design procedure on two-phase flow thermal-hydraulics of light water-cooled reactors. In order to perform thermal hydraulic analysis of SCWR, ACE-3D is enhanced to supercritical pressure region. As a result, it is confirmed that transient change in subcritical and supercritical pressure region can be simulated smoothly using ACE-3D, that ACE-3D can predict the results of the past heat transfer experiment in the supercritical pressure condition, and that introduction of thermal conductivity effect of the wall restrains fluctuation of wall temperature.
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  • Hiroyuki YOSHIDA, Takeharu MISAWA, Kazuyuki TAKASE
    2009 Volume 3 Issue 1 Pages 204-215
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method or particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D that adopts boundary fitted coordinate system in order to simulate complex shape flow channel. In this paper, boiling two-phase flow analysis in a tight-lattice rod bundle was performed by ACE-3D code. The parallel computation using 126 CPUs was applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. The tendency of void fraction distribution agreed with the measurement results by neutron radiography qualitatively. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight-lattice rod bundle with no lift force model was also performed. From the comparison of calculated results, it was concluded that the effects of lift force model were not so large for overall void fraction distribution of tight-lattice rod bundle. However, the lift force model is important for local void fraction distribution of fuel bundles.
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  • Yoshio HONJO, Masahiro FURUYA, Tomoji TAKAMASA, Koji OKAMOTO
    2009 Volume 3 Issue 1 Pages 216-227
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation-induced surface activation (RISA) hydrophilicity. In order to investigate gamma ray-induced and photoinduced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to the irradiation time before the contact angles reached its super-hydrophilic state. The irradiation time dependency is equal to each other qualitatively. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly indicates that certain humidity is required in ambient gas to take place of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the O1s spectrum, which indicates the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface.
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  • Jun ARAI, Seiichi KOSHIZUKA
    2009 Volume 3 Issue 1 Pages 228-236
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    We investigated single droplet impingement to a solid wall by the MPS-AS method. The MPS-AS method was a unified particle method based on the MPS method for both compressible and incompressible flows. The speed and the profile of the shock wave from the impact point agreed well with the experimental result by Field et al. The shock reflection from the droplet surface was also captured. The same tendency as Rochester's experimental approximation was obtained with respect to the peak pressure.
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  • Masa-aki TANAKA, Hiroyuki OHSHIMA, Hideaki MONJI
    2009 Volume 3 Issue 1 Pages 237-248
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    At the Japan Atomic Energy Agency (JAEA), the 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 the turbulence mixing of fluids at different temperatures. In this paper, numerical schemes for thermal-hydraulic simulation employed in MUGTHES are described, including the LES model. A simple method to limit numerical oscillation is adopted in energy equation solutions. A new iterative method to solve the Poisson equation in the BFC system is developed for effective transient calculations. This method is based on the BiCGSTAB method and the SOR technique. As the code validation of MUGTHES, a numerical simulation in a T-junction piping system with the LES approach was conducted. Numerical results related to velocity and fluid temperature distributions were compared with existing water experimental data and the applicability of numerical schemes with the LES model in MUGTHES to the thermal striping phenomenon was confirmed.
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  • Hidemasa YAMANO, Yoshiharu TOBITA
    2009 Volume 3 Issue 1 Pages 249-260
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: molten fuel freezing and dispersion; and boiling behavior of molten fuel pool. Related experimental database are reviewed to select appropriate experiments. To analyze the fuel freezing behavior, the GEYSER out-of-pile and the CABRI-EFM1 in-pile experiments were selected. The SIMMER-III calculations were in good agreement with fuel penetration lengths measured in a series of the GEYSER experiments. The fuel freezing behavior in the CABRI-EFM1 experiment was also reasonably simulated by SIMMER-III. The boiling pool consisting principally of molten fuel/steel mixtures is characterized by the heat transfer between fuel and steel. The CABRI-TPA2 experiment has suggested low transient heat flux from fuel to steel due to a steel vapor blanketing around a steel droplet. SIMMER-III well simulated the steel boiling behavior observed in the CABRI-TPA2 experiment by applying reduced heat transfer between fuel and steel.
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  • Takahiro ITO
    2009 Volume 3 Issue 1 Pages 261-271
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    This study is made to elucidate the primary mechanism dominating the inception and progress of eutectic melting between two solid metals contacting with each other. In this study, Cu-Ag eutectic system is simulated with classical molecular dynamics using Embedded Atom Method(1). First, melting temperature of solid solution of Cu-Ag binary system is investigated. The melting temperature depends on the atomic concentration of Cu and follows the liquidus line obtained in the experiments. The minimum melting temperature was obtained at the eutectic concentration. The melting behavior on the interface between two pure Cu and Ag slabs are then simulated. The mutual diffusion at the interface was considerably enhanced by the surface melting of both the metals. It is shown that the melting temperature at the interface is lowered depending on the local value of the atomic fraction and is almost identical to that of the solid solution with the corresponding atomic fraction.
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  • Makoto SHIBAHARA, Qiusheng LIU, Katsuya FUKUDA
    2009 Volume 3 Issue 1 Pages 272-288
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    Forced convection transient heat transfer for helium gas at various periods of exponentially increasing heat input (Q0exp(t/τ)) to a horizontal plate (ribbon) was experimentally and theoretically studied. In the experimental studies, the authors measured heat flux, surface temperature, and transient heat transfer coefficients for forced convection flow of helium gas over the horizontal plate under wide experimental conditions. The platinum plate with a length of 50 mm was used as a test heater. The gas flow velocities ranged from 4 to 10 m/s, the gas temperatures ranged from 313 to 353 K, and the periods of heat generation rate, τ, ranged from 46 ms to 17 s. The pressures were from 400 to 800 kPa. It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period longer than about 1 s, and it becomes higher for the period shorter than around 1 s. Empirical correlations for quasi-steady state heat transfer and transient one were obtained based on the experimental data under various pressures. In the theoretical study, transient heat transfer was numerically solved based on a turbulent flow model. The values of numerical solutions for surface temperature and heat flux were compared and discussed with authors' experimental values. It was obtained that the surface temperature difference and heat flux increase exponentially as the heat generation rate increases with the exponential function. It is understood that the gradient of the temperature distribution near the heater surface is higher at a higher surface temperature difference. The values of numerical solutions for heat flux agree well with the experimental data, though surface temperatures show some differences.
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  • Akira FUKUICHI, Yutaka ABE, Akiko FUJIWARA, Yujiro KAWAMOTO, Chikako I ...
    2009 Volume 3 Issue 1 Pages 289-300
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    One of the most interesting devices for light water reactor systems aimed at simplified system, improvement of safety and reliability is a supersonic steam injector. Supersonic steam injector is a passive jet pump without rotating machine and high efficient heat exchanger because of direct contact condensation between supersonic steam and a subcooled water jet. It is considered that flow behavior in the supersonic steam injector is related to complicated turbulent flow with large shear stress induced by velocity difference between steam and water and direct contact condensation. However, studies about turbulent flow under large shear stress with direct contact condensation are not enough. Especially, mechanisms of momentum and heat transfer are not clarified in detail. Objective of the present study is to investigate turbulent behaviors of a water jet and interface that play an important role in heat transfer and momentum transfer. Radial distribution of streamwise velocity and fluctuation of total pressure are measured by a pitot measurement. Visual measurement of the turbulent water jet is conducted by a high speed camera in order to identify location of unstable interface and its behavior. It is found that streamwise velocity increases as it approaches downstream of the mixing nozzle. Fluctuation of total pressure is large at water-steam mixture region. It is confirmed that waves propagated on the interface. And its velocity is obtained.
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  • Wei LIU, Masatoshi KURETA, Kazuyuki TAKASE
    2009 Volume 3 Issue 1 Pages 301-312
    Published: 2009
    Released: February 27, 2009
    JOURNALS FREE ACCESS
    This paper concerns experimental research to ascertain the effect of axial power distribution on critical power in the positive quality region. Experiments took place at atmospheric pressure in a circular tube. Axial uniform heating and two other axial non - uniform heating cases were selected for detailed evaluation. The effects of relative power ratio on critical power, critical quality and critical boiling length were ascertained in detailed evaluations. Using the experimental data, we evaluated existing correlating concepts with critical power. Result showed a combination of the overall power concept (χBT - LB) and the local conditions concept (χBT - qBT) appearing to be promising in correlating present critical power data in axial non - uniform heating conditions.
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  • Shuai ZHANG, Koji MORITA, Noriyuki SHIRAKAWA, Yuichi YAMAMOTO
    2009 Volume 3 Issue 1 Pages 313-320
    Published: 2009
    Released: March 11, 2009
    JOURNALS FREE ACCESS
    The COMPASS code is a new next generation safety analysis code to provide local information for various key phenomena in core disruptive accidents of sodium-cooled fast reactors, which is based on the moving particle semi-implicit (MPS) method. In this study, improvement of basic fluid dynamics models for the COMPASS code was carried out and verified with fundamental verification calculations. A fully implicit pressure solution algorithm was introduced to improve the numerical stability of MPS simulations. With a newly developed free surface model, numerical difficulty caused by poor pressure solutions is overcome by involving free surface particles in the pressure Poisson equation. In addition, applicability of the MPS method to interactions between fluid and multi-solid bodies was investigated in comparison with dam-break experiments with solid balls. It was found that the PISO algorithm and free surface model makes simulation with the passively moving solid model stable numerically. The characteristic behavior of solid balls was successfully reproduced by the present numerical simulations.
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