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Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, J.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/10
Suzuki, Hideaki; Nakama, Shigeo; Kimura, Makoto; Fujita, Tomoo
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/09
Yamano, Hidemasa; Tanaka, Masaaki; Kimura, Nobuyuki; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 18 Pages, 2009/09
This paper describes the current status of flow-induced vibration evaluation methodology development for the primary cooling pipings in JSFR, in particular emphasizing on R&D activities that investigate unsteady hydraulic characteristics in a short-elbow piping. Experimental efforts have been made using 1/3-scale and 1/10-scale single-elbow test sections for the hot-leg piping and 1/4-scale and 1/7-scale triple-elbow test sections for the cold-leg piping. Simulation activities include Unsteady Reynolds Averaged Naviar Stokes equation (U-RANS) approach with Reynolds Stress Model (RSM) using a CFD code and Large Eddy Simulation (LES) approach using in-house codes. Numerical results appears in this paper, focusing on its applicability to the hot-leg piping experiments. The numerical results could be provided to the input data for the structural vibration evaluation of the piping. The procedure of the flow-induced vibration evaluation is also described in this paper.
Yuki, Kazuhisa*; Hasegawa, Shunsuke*; Sato, Tsukasa*; Hashizume, Hidetoshi*; Aizawa, Kosuke; Yamano, Hidemasa
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 11 Pages, 2009/09
In this study, the flow structures in a 3-dimentionally connected dual elbow, which simulates a part of the cold leg, are visualized by PIV measurement, and it is discussed on the detailed flow transition from the 1st elbow to the 2nd elbow and the generation of unsteady flow such as a separation that influences flow-induced vibration. Experimental apparatus has a 1/15 scale test section of actual design whose inner diameter and curvature radius ratio are 56mm and 1.0, respectively. For visualization without any image distortion, matched refractive-index PIV measurement is carried out using NaI solution as the working fluid. The Reynolds number is 50,000, and the inlet flow condition to the test section is a fully developed turbulent flow. It is confirmed that it generates a separation along the inner wall of the 1st elbow and one large swirling flow in the 2nd elbow. Furthermore, unsteady flow formed in and/or behind the separation region is transported downstream and flows into the center area of the 2nd elbow.
-number short-elbow pipe flowEguchi, Yuzuru*; Murakami, Takahiro*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 14 Pages, 2009/09
Since unsteady force possibly generated by fluid motion in a short-elbow pipe needs evaluating in the JSFR plant design, applicability of large-eddy simulation (LES) method to a high number water test is studied with an in-house LES code in the paper. First, requirement for mesh subdivision at high- LES computation is theoretically studied. Then, the effects of turbulence model and boundary conditions are numerically studied using the Smagorinsky model and the WALE (Wall-adapting local eddy-viscosity) model with various inlet and outlet boundary conditions at = 1.2
10
. It has turned out that the Smagorinsky model excels the WALE model and that an inlet velocity profile has a considerable impact on the separation features at the elbow curvature. It was also revealed that the outlet boundary condition is likely to have an effect on the separation features especially if the distance between the elbow and the outlet is rather short, which is the case of the present computation. Basic features of the pressure and the wall force fluctuations are also discussed with the aid of computational visualization and spectral analysis.
Suzuki, Mitsuhiro; Nakamura, Hideo
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 17 Pages, 2009/09
Presented are experiment results on performance of core exit thermocouple (CET) and applicability to PWR accident management (AM) during 12 tests of small-break loss-of-coolant accident (SBLOCA) and abnormal transient conducted at the Large Scale Test Facility (LSTF) of Japan Atomic Energy Agency, which is the largest PWR simulator with full-height and 1/48 volume scaling. General CET performances are derived including (1) CETs are capable in most cases to detect core overheating with delay of time and temperature increase from core heat-up, (2) one of the reasons of this delay is attributed to cooling effects of structural materials at the core exit and peripheral region, (3) CETs were incapable to detect core overheating in a very small break under steam generator depressurization action as well as a 10% cold leg break due to significant water fall-back from hot legs, and (4) an alternative indication by CET superheat is necessary in extremely high and low pressure conditions.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Inagaki, Yoshiyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09
Xenon spatial oscillations on the cylindrical cores used in light water reactors such as BWR and PWR have already been theoretically and experimentally much-studied. On the other hand, evaluations of the oscillations on annular cores applied to VHTR have still not been fully evaluated. It is essential to investigate the relationship among the oscillations, annular core geometry and feedback effect with temperature coefficient. In this paper, a neutron binding parameter related to core geometry as one of dimensionless parameters are proposed and these parameters are useful for the evaluation of xenon spatial oscillations in annular cores.
Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 11 Pages, 2009/09
Kimura, Nobuyuki; Ono, Ayako; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 16 Pages, 2009/09
A quantitative evaluation on high cycle thermal fatigue due to temperature fluctuation in fluid is of importance for structural integrity in the reactor. It is necessary for the quantitative evaluation to investigate occurrence and propagate processes of temperature fluctuation, e. g. decay of temperature fluctuation near structures and transfer of temperature fluctuation from fluid to structures. In this study, a water experiment using T-junction was performed to evaluate the transfer characteristics of temperature fluctuation from fluid to structure. In the experiment, temperatures and local velocities were measured simultaneously to evaluate the correlation between the temperature and velocity under the non-stationary fields. The large heat transfer coefficients were registered at the region where the local velocity was high.
Ohno, Shuji; Ohshima, Hiroyuki; Sugahara, Akihiro*; Oki, Hiroshi*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09
Multi-dimensional transient thermal hydraulic analyses for a sodium experiment are carried out in order to validate the applicability of basic simulation methods to a typical thermal stratification behavior in a reactor upper plenum region. Simulation predictability of the behavior is evaluated after the results of an existing experiment are reviewed and important fundamental characteristics of the behavior are quantified. The investigated results clarify that CFD codes provide good prediction for fundamental phenomena of steep temperature gradient and gradual rising behavior of the stratification interface under the condition of adopting appropriate mesh size, higher-order discretization scheme, and the RANS turbulence model.
Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 16 Pages, 2009/09
According to the compact sizing of reactor vessel, gas entrainment at the free surface of sodium coolant becomes one of the significant issues in the latest sodium-cooled fast reactor design. In the present study, some basic experiments for gas entrainment due to surface vortices were performed in order to obtain the fundamental knowledge about entrained bubbles. The distribution of bubble diameters was obtained from an image processing technique from the images. In addition to that, velocity field and surface shape of gas core were measured to grasp the influences of the local velocity field on the gas core shape and detached bubbles. The results showed that the mean equivalent diameters of bubbles were varied from 1.34 to 2.06 mm in the range of present experimental conditions.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 14 Pages, 2009/09
The main cooling system of Japan Sodium-Cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9 m/s in the primary hot leg pipe having a short elbow with large-diameter. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned. The evaluation of the structural integrity of pipings in JSFR will need to grasp the mechanism of FIV at the elbow. In this study, two types of 1/8 scaled elbows with different curvature ratio were used to investigate the influence of curvature on velocity fluctuation. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow and separation flow in the elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow.
Kamide, Hideki; Kobayashi, Jun; Hayashi, Kenji
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 17 Pages, 2009/09
Natural circulation has a significant role in the decay heat removal function of a sodium cooled reactor. The dipped heat exchanger (DHX) immersed in a reactor upper plenum as the decay heat removal system provides cold sodium in the upper plenum during the decay heat removal operation. This cold sodium covers the top of the core under low flow rate conditions of the natural circulation. Sodium experiments were carried out to find onset condition and penetration depth of such partial reverse flow driven by buoyancy force. The onset conditions of the penetration flow were correlated with Gr and Re numbers as well as basic water experiments. The observed penetration depths were limited in the upper axial neutron shielding of the subassembly.
Ito, Kei; Kunugi, Tomoaki*; Ohshima, Hiroyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/09
For the purpose of direct numerical simulations of gas entrainment (GE) phenomenon in sodium-cooled fast reactors, a high-precision numerical simulation method for gas-liquid two-phase flows have been developed. To reproduce the GE phenomenon, the locality of the GE phenomenon has been addressed by developing an unstructured adaptive mesh technique by which high mesh resolutions can automatically applied transiently to the occurrence regions of the GE phenomenon. In this paper, the unstructured adaptive mesh technique was extended to two-phase flows. For that purpose, the redistribution algorithms for two-phase flow variables were developed based on the conservation laws of the variables. The unstructured adaptive mesh technique was validated by solving the well-known slotted-disk revolution and dam-break problems. As a result, the adaptive mesh technique succeeds in maintaining the slotted-disk shape after one revolution and resolving the dynamic dam-breaking behavior.
Onoda, Yuichi; Fukano, Yoshitaka; Sato, Ikken; Marquie, C.*; Duc, B.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 15 Pages, 2009/09
Matsumoto, Toshinori*; Takata, Takashi*; Yamaguchi, Akira*; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 10 Pages, 2009/09
When a heat transfer tube fails in steam generator (SG) of sodium-cooled fast reactor (SFR), sodium-water reaction (SWR) would take place. It is significant for estimation of the heat transfer from the fluid to the tube wall during SWR region to evaluate the possibility of the secondary tube failure in case of overheating rupture. In the present study, thermal hydraulics simulation of the fluid around the tube is conducted. The heat transfer coefficient is computed, the correlation diagram between the heat transfer coefficient and the void fraction has been obtained. The void fraction around the heat transfer tube in the SWR has been evaluated.
Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/09
A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. It is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle for the high burn-up core. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The velocity distribution in an inner subchannel with the wrapping wire was measured by Particle Image Velocimetry. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetrical flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed.
Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09
In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding/neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding/n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R&Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics.
Garin, P.*; Sugimoto, Masayoshi
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 11 Pages, 2009/09
In June 2007 were launched the Engineering Validation and Engineering Design Activities of the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) in the framework of the Broader Approach Agreement between Euratom and Japan. The main objective of the project is the delivery of the Engineering Design Report of IFMIF, enabling its future construction. The main technologies shall be demonstrated through the design, the construction and the experiment of three prototypes: the Accelerator, the Lithium Target and the High Flux Test Module of the Test Facilities. Since the beginning of the project, focus has been put on the prototypes definition and 2009 will be the actual start of IFMIF engineering design activities. These activities will be described in the paper, as well as the status of the prototypes, which have considerably evolved since the former conceptual phase of IFMIF.
Liu, W.; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 10 Pages, 2009/09
For the Steam Generator (SG) in a commercialized sodium cooled faster breeder reactor, flow instability in water side is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments using water as test fluid were performed under high pressure condition at JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper focuses on the discussion to steam - water pressure drop. We evaluated existing correlations for two-phase flow multiplier under high pressure. As a result, Chismholm correlation was confirmed being the best one for the present high pressure data.