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Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.15 - 16, 2011/06
For the purpose of elucidating the mechanism of the sodium-water reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting point of Na and NaOH, transition temperature of NaOH, Na-NaOH reaction temperature, and the decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of Na
O generation was obtained. Thermal analysis results indicated that NaO generation at the secondary overall reaction would be considered during the sodium-water reaction.
Liu, W.; Takase, Kazuyuki
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.395 - 398, 2011/06
For a steam generator with straight double-walled heat transfer tubes that will be used to a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. As the first step of the research, thermal hydraulics experiments with water were performed under the high pressure condition with using a circular tube at JAEA. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the heat transfer characteristics under 15-18 MPa. Saturated boiling heat transfer was discussed with four most general heat transfer correlations (Chen, Shah, Steiner-Taborek and Gungor-Winterton). Under the present high pressure condition, it was found that the Shah correlation gave a good agreement with data at low mass flow rate and the Chen correlation gave a good agreement at high mass flow rate condition. For the nominate flow rate of w =110 g/s, both correlations of Chen and Shah can be used. As a result, under the present high pressure condition, we recommend that the smaller one of the Chen and Shah correlations can be used for the calculation of heat transfer coefficient.
Hiyama, Tomoyuki; Doda, Norihiro; Ohshima, Hiroyuki; Iwasaki, Takashi*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.217 - 218, 2011/06
An analysis of JOYO MK-II natural circulation tests has been performed to evaluate the applicability of the fast reactor plant dynamic analysis code, Super-COPD, to natural circulation decay heat removal phenomena. In this analysis, the predicted transient behaviors of the core outlet coolant temperature and the primary flow rate showed good agreement with the test results by using a variable mesh partitioning method for the upper plenum region which can include the effect of thermal stratification phenomena.
Kurihara, Akikazu; Umeda, Ryota; Yanagisawa, Hideki*; Ohshima, Hiroyuki
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.9 - 10, 2011/06
In the case of sodium-water reaction accident in a steam generator of sodium-cooled fast reactors (FRs), adjacent heat transfer tubes may be damaged due to high temperature environment of the reaction field. For the purpose of understanding the overheating tube rupture mechanism, an experimental study has been performed to clarify waterside heat transfer characteristics during up-flow in a vertical tube under the real plant part-load operation conditions in which safety margin is least. A test tube was heated rapidly and the time averaged heat flux was estimated using an inverse solution. It was conformed that the heat transfer on the wall changed from nucleate boiling to transient-film boiling all over the heating section and dried-out surface spread from downstream toward upstream. We improved the heat transfer correlations applied to RELAP5 code and made sure the adequacy of these correlations to evaluate tube overheating.
Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.109 - 110, 2011/06
no abstracts in English
Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki; Ikuta, Ryuhei*; Koizumi, Yasuo*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.103 - 106, 2011/06
no abstracts in English
Ohshima, Hiroyuki; Yamaguchi, Akira*; Narabayashi, Tadashi*; Deguchi, Yoshihiro*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.1 - 2, 2011/06
When a heat transfer tube is failed in a steam generator (SG) of a sodium-cooled fast reactor (SFR), pressurized water and/or water vapor leaks into liquid sodium surrounding the tube and forms a reacting jet with high temperature. This reacting jet might cause the secondary failure of adjacent heat transfer tubes due to wastage or over-heating tube rapture resulting in undesirable development of the accident. Therefore, the sodium-water reaction phenomenon (SWR) is one of most important issues for the design and safety assessment of SFRs. This paper describes the research and development plan of a new multi-physics numerical simulation system which is based on mechanistic and theoretical modeling of the SWR rather than empirical modeling and can contribute to detailed and quantitative evaluations of the SWR in any types of SGs including commercial SFRs.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.215 - 216, 2011/06
In the design study for Japan Sodium Cooled Fast Reactor (JSFR), fully natural circulation system is adopted as the decay heat removal system. We have been developing a new evaluation method of core hot spot in transition from rated operation to natural circulation decay heat removal conditions. Since the method is currently based on conservative assumptions and data, there is room for further rationalization of the safety margin which can be achieved by conducting best estimate analyses with confidence and with quantified uncertainty of results. This paper describes a development of PIRT (Phenomena Identification and Ranking Table) for JSFR under natural circulation decay heat removal operations and the sensitivity analyses of the uncertainties in the event of loss of external power as the first step to improve the evaluation method.
Ito, Kei; Koizumi, Yasuo*; Ohshima, Hiroyuki; Kawamura, Takumi*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.219 - 222, 2011/06
Gas entrainment (GE) phenomena in fast reactors have to be evaluated in terms of their occurrences because entrained bubbles in the primary cooling system may cause disturbances in reactor power. In fast reactors, a small amount of entrained gas may be allowed because there are always small deposition bubbles in the primary cooling system. In this case, not only the GE occurrences but also the amount of the entrained gas should be evaluated accurately. Therefore, the authors are studying the amount of the entrained gas by utilizing simple an experiment and a numerical simulation. In this paper, parametric simulations of the simple experiment are performed with a high-precision volume-of-fluid algorithm which simulate interfacial dynamic deformations directly. As a result, the simulation results gives larger entrained gas flow rate than the experimental data but the dependency of the entrained gas flow rate on the liquid flow rate is reproduced qualitatively.
Tanaka, Masaaki; Takita, Hiroki*; Monji, Hideaki*; Ohshima, Hiroyuki
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.223 - 224, 2011/06
Water experiment and numerical simulation under thermal interaction conditions between fluid and structure were conducted for a T-junction piping system (T-pipe) consisting of a rectangular duct for main stream and a circular pipe for branch stream. Numerical results were compared with the experimental results and indicated that the fluid-structure thermal interaction was necessarily considered for thermal fatigue estimation in the thermal striping phenomena because fluid temperature distribution near the wall was much affected by the thermal interaction.