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Kawasaki, Nobuchika; Nagae, Yuji; Kato, Shoichi; Ando, Masanori; Kasahara, Naoto
Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.507 - 510, 2008/06
Concept of development in strength evaluation for fast reactor vessels was introduced. Based on the characteristics of reactor vessel for developing fast breeder reactor, creep-fatigue strength and ratcheting deformation criterion are under developing. For the creep-fatigue strength, intermediate dwelling creep-fatigue, aging effect and strain concentration are selected as developing items. The ratcheting deformation criterion will be determined by ratcheting fatigue and ratcheting creep-fatigue test results.
Sakamoto, Yoshihiko; Kubo, Shigenobu*; Kotake, Shoji; Kamishima, Yoshio*
Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.505 - 506, 2008/06
This paper describes the enhancement of reliability for components of the reactor system in JSFR design. As for manufacturability, compact design of the RV enables its manufacture in a factory. This results in high quality welding and in precise machining of the RV. The adoption of ring-shaped forgings contributes for securing the reliability against thermal stress as well as securing the dimension precision. Regarding maintenability, the in-vessel structures have simple configurations, so it is comparatively easy for inspection equipments to reach inspection targets. In the JSFR design, sodium boundary area is reduced significantly, which makes double-walled design of the piping easier, and reduces welding lines. So, the reactor system of JSFR is desirable to inspect the in-vessel structures efficiently, and there is a prospect of reliable plant operation. Advanced inspection technologies are also under development for the inspection of the in-vessel structures under sodium.
Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki*
Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.495 - 496, 2008/06
In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of commercialized sodium-cooled fast breeder reactor. In this study, the computer code for flow instability analysis in single heat transfer tube was developed with drift-flux model which included the effects of subcooled boiling and two phase slip. The special algorithm to calculate inlet flow rate with inlet pressure, outlet pressure and heat flux as boundary conditions for the density-wave instability analysis has been established. The subcooled model was verified by calculating the void fraction distribution of steady heat transfer flow. The capability of drift flux model for simulating density-wave instability in single tube was confirmed.
Ohno, Shuji; Oki, Hiroshi*; Sugahara, Akihiro*; Ohshima, Hiroyuki
no journal, ,
Three-dimensional thermal-hydraulic analyses of thermal stratification phenomena have been carried out in order to validate simulating ability of single-phase thermal hydraulic code AQUA and commercial CFD codes STAR-CD and FLUENT. The analyses of thermal stratification water experiments demonstrated that the codes reproduced temperature gradient and upward movement of the stratification interface in the case of utilizing appropriate discretization method and computational mesh arrangement for gravitational direction. No remarkable difference was observed between the calculated results with three codes. It was also shown that three turbulence models of standard k-
, RNG k- and RSM indicated little difference in the predicted results of basic stratification phenomena observed in the experiments.
Watanabe, Hironori; Mitsutake, Toru*; Kakizaki, Sadayuki*; Sekine, Katsunori*; Takase, Kazuyuki
no journal, ,
no abstracts in English
Ito, Kei; Kunugi, Tomoaki*; Ohshima, Hiroyuki
no journal, ,
To evaluate directly gas entrainment (GE) phenomena in fast breeder reactors, we have been studied a numerical simulation method based on a high-precision volume-of-fluid methodology. We have been employed unstructured meshes because exact modeling of complicated geometries in each domain is a key to simulate GE phenomena accurately. Therefore, formulations of each calculation procedure in the high-precision VOF methodology on unstructured meshes are conducted in this paper. Then, unphysical behaviors of velocity distributions near gas-liquid interface induced by inappropriate formulation of pressure gradient are discussed and an appropriate formulation is derived. Finally, the basic GE experiment is simulated using our numerical simulation method. The simulation results shows that the GE phenomena occurs in the same mechanism with the experimental results.
Uchibori, Akihiro; Ohshima, Hiroyuki
no journal, ,
A numerical analysis method for melting/solidification phenomena has been developed to evaluate feasibility of the several candidate techniques in the nuclear fuel cycle. Our method is based on the eXtended Finite Element Method (X-FEM) which has been used for moving boundary problems. The basic idea of the X-FEM is to incorporate the signed distance function into the standard finite element interpolation to represent a discontinuous gradient of the temperature at a moving solid-liquid interface. This technique makes it possible to simulate movement of the solid-liquid interface without the use of a moving mesh. The numerical solutions of the basic problems, a one-dimensional Stefan problem, solidification in a two-dimensional square corner and melting of pure gallium, were compared to the exact solutions or to the experimental data. Through these verifications, validity of the newly developed numerical analysis method has been demonstrated.