Thermal-Hydraulic Research in View of Nuclear Safety

The accident at Fukushima Dai-ichi Nuclear Power Station of Tokyo Electric Power Company entirely lost people\u27s trust in nuclear power. We learned from the accident the importance of continuous effort to pay attention to improving the safety of nuclear power plant by reflecting new knowledge and experience. Concerning thermal-hydraulics research, severe accident research in particular, the question arises whether or not research results were reflected to safety measures or those obtained deserved it. In view of these, thermal-hydraulics research concerning nuclear safety up to the present will be reviewed and some suggestions will be presented in this paper

ICONE10-22746 RADIATION INDUCED SURFACE ACTIVITY PHENOMENON (2nd report: Radiation Induced Boiling Enhancement)

ABSTRACT To delineate the effect of Radiation Induced Surface Activity (RISA) on boiling phenomenon, surface wettability in high-temperature environment or Leidenfrost condition and critical heat flux (CHF) of oxide metals irradiated by gamma rays were investigated. When the temperature of the heating surface reaches the wetting limit temperature, water-solid contact vanishes because of a stable vapor film between the droplet and the metal surface, i.e., a Leidenfrost condition. The wetting limit temperature increased with integrated irradiation dose. The CHF of oxidized titanium was improved up to 100% after 800 kGy 60 Co gamma ray irradiated. Radiation Induced Boiling Enhancement (RIBE) phenomenon was firstly confirmed through the experiments

低圧力, 低流量におけるバーンアウトに関する研究

京都大学0048新制・論文博士工学博士乙第5303号論工博第1695号新制||工||602(附属図書館)UT51-59-F367(主査)教授 西原 英晃, 教授 岐美 格, 教授 櫻井 彰学位規則第5条第2項該当Kyoto UniversityDFA

Shape measurement of bubble in a liquid metal

Dynamic behavior of a two-phase bubble, i.e. a steam bubble containing a droplet evaporating in the bubble, in the molten alloy was clearly visualized using high-frame-rate neutron radiography. In relation to some direct contact heat exchanger design with molten lead-bismuth (Pb-Bi), experiments have been done at JRR-3M of JAEA (Japan Atomic Energy Agency) with water droplets evaporating in a stable thermally stratified Newton's alloy pool. The instantaneous shape and size of the bubble has been iteratively estimated from the void fraction distributions and total void volume by assuming a symmetrical bubble shape

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2 gU/cm3) and enrichment (20%) of LEU fuel (U3Si2–AL) are quite different from the uranium density (0.58 gU/cm3) and enrichment (93%) of HEU fuel (U–Al), which may result in the changes of heat transfer response and neutronic characteristic in the core. So it is necessary to objectively re-assess the feasibility of LEU silicide fuel core in KUR by using various numerical simulation codes. This paper established a detailed simulation model for the LEU silicide core and provided the safety analyses for the reactivity insertion transients in the core by using EUREKA-2/RR code. Although the EUREKA-2/RR code is a proven and trusted code, its validity was further confirmed by the comparison with the predictions from another two thermal hydraulic codes, COOLOD-N2 and THYDE-W at steady state operation. The steady state simulation also verified the feasibility of KUR to be operated at rated thermal power of 5 MW. In view of the core loading patterns, the operational conditions and characteristics of the reactor protection system in KUR, the accidental control rod withdrawal transients at natural circulation and forced circulation modes, the cold water injection induced reactivity insertion transient and the reactivity insertion transient due to removal of irradiation samples were conservatively analyzed and their transient characteristic parameters such as core power, fuel temperature, cladding temperature, primary coolant temperature and departure from nucleate boiling ratio (DNBR) due to the different ways and magnitudes of reactivity insertions were focused in this study. The analytical results indicate that the quick power excursions initiated by the reactivity insertion can be safely suppressed by the reactor protection system of KUR in various initial power levels and different operational modes (natural circulation and forced circulation modes). No boiling and no burnout on fuel cladding surface and no blister in the fuel meat happens and KUR is safe in all of these reactivity insertion transients if the reactor protection system of KUR works in its minimum degree

Status of prediction methods for critical heat fluxes in mini and microchannels

Saturated critical heat flux (CHF) is an important issue during flow boiling in mini and microchannels. To determine the best prediction method available in the literature, 2996 data points from 19 different laboratories have been collected since 1958. The database includes nine different fluids (R-134a, R-245fa, R-236fa, R-123, R-32, R-113, nitrogen, CO2 and water) for a wide range of experimental conditions. This database has been compared to 6 different correlations and I theoretically based model. For predicting the non-aqueous fluids, the theoretical model by Revellin and Thome [Revellin, R., Thome, J.R., 2008. A theoretical model for the prediction of the critical heat flux in heated microchannels. Int. J. Heat Mass Transfer 51, 1216-1225] is the best method. It predicts 86% of the CHF data for non-aqueous fluids within a 30% error band. The data for water are best predicted by the correlation by Zhang et al. [Zhang, W., Hibiki, T., Mishima, K., Mi, Y., 2006. Correlation of critical heat flux for flow boiling of water in minichannels. Int. J. Heat Mass Transfer 49, 1058-1072]. This method predicts 83% of the CHF data for water within a 30% error band. Some suggestions have also been proposed in this paper for the future studies. (C) 2009 Elsevier Inc. All rights reserved