Effect of the rapid evaporation on the motion of melt drops during the coarse mixing process of vapor explosions

Y.Yang*; *; Sugimoto, Jun

Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 2, p.663 - 670, 1997/00

no abstracts in English

Conceptual design study of IFMIF target system

; Nakamura, Hideo; *; Maekawa, Hiroshi; Katsuta, Hiroji; T.Hua*; S.Cevolani*

Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 3, p.1260 - 1267, 1997/00

no abstracts in English

Heat Removal Characteristics from a 36-rod Fuel Bundle in a Tube by Radiative Heat Transfer during LOCAs without Emergency Coolant Injection

Mochizuki, Hiroyasu;

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 0 Pages, 1997/00

None

Numerical Analyses of Self-Induced Free Surface Flow Oscillation by Fluid Dynamics Computer Code SPLASH-ALE

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 0 Pages, 1997/00

None

None

Sakai, Takaaki

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 0 Pages, 1997/00

None

Thermal Fuid-Structure Interaction Analysis for the Upper Structure of LMFRs

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.2, p.856 - 863, 1997/00

None

Improvement of CHF correlations for research reactors using plate-type fuels

Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1815 - 1822, 1997/00

In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as "Loss of the primary coolant flow". Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and nonuniform heat flux condition. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed.

Numerical analysis of thermal stratification phenomean in upper Plenum of fast breeder reactor

;

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, 0 Pages, 1997/00

None

SIMMER-III Applications to Key Phenomena of CDAs in LMFR

Morita, Koji; Tobita, Yoshiharu; ;

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1332 - 1339, 1997/00

None

Conceptual Design of a Potassium Turbine System for Transportable Reactor

; ; Seino, Hiroshi; Kataoka, H.

p1101-1110, p.1101 - 1110, 1997/00

None