Thermal fatigue experiment of screw cooling tube under one-sided heating condition

Ezato, Koichiro; Suzuki, Satoshi; Sato, Kazuyoshi; Akiba, Masato

Journal of Nuclear Materials, 329-333(1), p.820 - 824, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.138

This paper presents thermal fatigue experiments of a cooling tube with a helical triangular fin on its inner cooled surface, namely a . The screw thread is directly shaped in a CuCrZr heat sink bar as a cooling channel. Slits with the width of 1.5 mm are machined at the heated side of the heat sink. The thermal fatigue experiments are carried out at 20 and 30 by using an electron beam irradiation facility in JAERI. Water leakages from fatigue cracks, which locate at the slit of the heat sink, occurred at around 4500th and 1400th cycles at 20 and 30 , respectively. These results show good agreement with lifetime predictions using Manson-Coffin's law based on finite element analyses. Fractographic observations reveal that the fatigue cracks start from the outer heated surface at the slit region of the cooling channel and propagate toward its inner surface.

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.

Experimental study of differences in CHF between upflow and downflow in vertical rectangular channels; Effect of subcooling

Kaminaga, Masanori; Sudo, Yukio

Nihon Kikai Gakkai Rombunshu, B, 58(553), p.2799 - 2804, 1993/09

https://doi.org/10.1299/kikaib.58.2799

no abstracts in English

A New CHF correlation scheme proposed for vertical rectangular channels heated from both sides in nuclear research reactors

Sudo, Yukio; Kaminaga, Masanori

J. Heat Transfer, 115, p.426 - 434, 1993/05

https://doi.org/10.1115/1.2910695

no abstracts in English

A Plotting system for three-dimensional oblique coordinates(triangular mesh) of CITATION

Kurosawa, Masayoshi

JAERI-M 92-007, 56 Pages, 1992/02

JAERI-M-92-007.pdf:1.86MB

no abstracts in English

Experimental study of the critical heat flux in a narrow vertical rectangular channel

Kaminaga, Masanori; Sudo, Yukio; Murayama, Yoji; *

Heat Transfer-Jpn. Res., 20(1), p.72 - 85, 1991/03

no abstracts in English

Experimental study of falling water limitation under a counter-current flow in a vertical rectangular channel, 1st report; Effect of flow channel configuration and introduction of CCFL correlation

Sudo, Yukio; *; Kaminaga, Masanori

JSME Int. J., Ser. 2, 34(2), p.169 - 174, 1991/00

no abstracts in English

Experimental results of angular neutron flux spectra leaking from slabs of fusion reactor candidate materials, I

Oyama, Yukio; ; Maekawa, Hiroshi

JAERI-M 90-092, 124 Pages, 1990/06

JAERI-M-90-092.pdf:4.94MB

no abstracts in English

Angular neutron flux measurements and nuclear data test on slabs of fusion blanket materials

Oyama, Yukio; Kosako, Kazuaki*; ; Maekawa, Hiroshi

Nuclear Data for Science and Technology, p.271 - 274, 1988/00

no abstracts in English

Spectral measurement of angular neutron flux on the restricted surface of slab assemblies by the time-of-flight method

;

Nuclear Instruments and Methods in Physics Research A, 245, p.173 - 181, 1986/00

https://doi.org/10.1016/0168-9002(86)90273-1

no abstracts in English

Experimental study of differences in DNB heat flux between upflow and downflow in a vertical channel

Sudo, Yukio; *; ; Kaminaga, Masanori;

Journal of Nuclear Science and Technology, 22(8), p.604 - 618, 1985/00

https://doi.org/10.3327/jnst.22.604

no abstracts in English