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Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 13 Pages, 2007/09
Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. This paper describes the validation results for the newly developed code using the experimental results of the safety demonstration tests. Especially, the reactivity was clarified using an original mathematical expression which shows the relationship among region temperature coefficient, region temperature rise and power distribution.
Kimura, Nobuyuki; Kamide, Hideki; Emonot, P.*; Nagasawa, Kazuyoshi*
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 14 Pages, 2007/09
A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing between hot and cold fluids causes thermal fatigue in structural components, is of importance for reactor safety. In this study, a water experiment of parallel triple jets configuration was performed in order to evaluate temperature fluctuation characteristics in fluid, transfer characteristics of temperature fluctuation from fluid to structure. The power spectrum densities of temperature fluctuation, furthermore, were in good agreements between the experiment and calculation. The calculated results showed that the heat transfer coefficient obtained from the experiment was appropriate and proper. The decay due to the heat transfer at the wall surface was dominant for the decay of temperature fluctuation at the neighborhood of the wall.
Ezure, Toshiki; Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 13 Pages, 2007/09
According to the compact sizing of reactor vessel, gas entrainment at the free surface of sodium coolant becomes one of the significant issues for the latest fast reactor design. It is required to clarify the criterion of gas entrainment at free surface and the tolerance of gas entrainment. In the present study, some visualization experiments were performed in the water-air system focusing on the gas entrainment due to a surface vortex. The gas entrainment occurs intermittently and the vortex develops and decays in time. Then developing process of the vortex is significant to estimate the gas entrainment phenomena. Multi plane PIV using a scanning mirror was adopted to investigate the temporal and spatial development of velocity around the vortex. The spatial propagation of downward velocity was clarified as the key factors of the vortex development.
Takeda, Takeshi; Asaka, Hideaki; Suzuki, Mitsuhiro; Nakamura, Hideo
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 3 Pages, 2007/09
RELAP5 code analysis was performed to validate the code predictability by using ROSA/LSTF experiment data that simulated a PWR vessel upper head small break loss-of-coolant accident (SBLOCA) with a break equivalent to 1% cold leg break. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient (Cd) of 0.61 for two-phase break flow. In the experiment, liquid level in the upper head was found to control break flow rate as coolant in the upper plenum entered the upper head through control rod guide tubes (CRGTs) until the penetration holes at the CRGT bottom were exposed to steam in the upper plenum. The upper head noding and flow paths between the upper plenum and the CRGT were thus modeled to simulate well the liquid level and coolant flow around the upper portion of pressure vessel. The code, however, overpredicted the break flow rate due to the underprediction of break-upstream void fraction especially during two-phase flow discharge period. Cd for two-phase break flow was thus adjusted to be 0.58. Effects of break area on the core cooling were investigated further. The parameter analyses showed that peak cladding temperature (PCT) is the maximum at 1% break case, while the PCT would be lower than 1200 K in the larger break size cases because vapor condensation on injected accumulator coolant induces loop seal clearing and effectively enhances core cooling thereafter.
Ito, Kei; Yamamoto, Yoshinobu*; Kunugi, Tomoaki*
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 15 Pages, 2007/09
For the purpose of direct numerical simulations of gas entrainment in fast breeder reactors, we are developing a high-precision seamless physical simulator based on computational scientific approaches. An unstructured mesh partitioning method was employed in this study and a high-precision calculation method for gas-liquid two-phase flow on the unstructured mesh was developed on the basis of the MARS formulated on a structured mesh. As the result of the verification, it was confirmed that the developed method has comparable or higher calculation accuracy compared to conventional methods. In addition, a correction method was introduced to the advection term of the volume fraction transport equation to improve volume conservation. The correction method led higher calculation accuracy on the unstructured mesh compared to the original method. Errors in a surface tension calculation were also estimated and the reconstructed distance function method showed the most accurate result.
Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 12 Pages, 2007/09
A breach of the heat transfer tube (HTT) in a steam generator (SG) of a Sodium Fast Reactor causes sodium-water reaction. The design and safety concern is a possibility of the secondary failure of nearby HTTs that could cause undesirable development of the accident. One needs to evaluate the temperature transients of the HTTs in the reaction region for safety evaluation. In the present study, a computational method is developed for this purpose. It solves the sodium thermal-hydraulics and the heat conduction in the adjacent HTTs. An experiment performed at JAEA is analyzed with the method developed in this study. It is found that analyzed temperatures are in good agreement with the experimental data. Based to the experimental and computational results, multi-phase multi-component flow characteristics are depicted. Furthermore, the heat transfer coefficient is evaluated using the numerical simulation result.
Takata, Takashi*; Yamaguchi, Akira*; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 12 Pages, 2007/09
A Large Eddy Simulation (LES) with a simple Sub-Grid Scale (SGS) model and a Near Wall Model (NWM) is one of the simplest ways to overcome the heavy computational cost. In the present paper, two kinds of benchmark analyses have been carried out in order to investigate an applicability of the simple LES with NWM to an engineering problem concerning with a nuclear plant thermal hydraulics; one is the temperature mixing in the parallel triple-jets and the other is the isothermal turbulent flow under a rod bundle configuration. In the analyses, the standard Smagorinsky model is chosen as a SGS model using a second order central differential scheme. As the NWM, a wall function and a boundary layer approximation are implemented. As a result of the benchmark analyses, it is demonstrated that one can investigate a temperature fluctuation which causes a thermal striping phenomenon and a flow pulsation in a rod bundle configuration with the present LES with NWM.
Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 12 Pages, 2007/09
Melting/solidification is a key phenomenon in the several nuclear fuel cycle processes. In this paper, numerical analysis of the basic problems of melting/solidification using an extended finite element method is presented. The method is based on an enriched finite element interpolation to represent a discontinuous gradient of the temperature at a moving solid-liquid interface. This enables us to simulate movement of solid-liquid interface without the use of a moving mesh. The numerical solutions of the basic problems, a one-dimensional Stefan problem, a problem of solidification in a two-dimensional square corner and a problem of melting of a pure gallium, are compared to the exact solutions or the experimental data. Through these verifications, it is demonstrated that the extended finite element method can be applied to melting/solidification phenomena.