Measurements of neutronic characteristics of rectangular and cylindrical coupled hydrogen moderators

Kai, Tetsuya; Kamiyama, Takashi*; Hiraga, Fujio*; Oi, Motoki; Hirota, Katsuya*; Kiyanagi, Yoshiaki*

Journal of Nuclear Science and Technology, 55(3), p.283 - 289, 2018/03

https://doi.org/10.1080/00223131.2017.1394235

Fabrication and test results of testing equipment for remote-handling of MA fuel, 2; Evaluation of heat transfer parameter crossing rectangular lattice matrix

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Inoue, Akira; Tsujimoto, Kazufumi

JAEA-Technology 2015-052, 34 Pages, 2016/03

JAEA-Technology-2015-052.pdf:5.02MB

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. So it is necessary to verify reliability and accuracy of heat transfer effect used in this area. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.

Numerical analysis of a water-vapor two-phase film flow in a narrow coolant channel with a three-dimensional rectangular rib

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada

JSME International Journal, Series B, 47(2), p.323 - 331, 2004/05

no abstracts in English

Experimental study on the critical heat flux (CHF) in a rectangular channel with micro ribs for a solid target and proton beam window design

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Sudo, Yukio

Proceedings of International Workshop on Current Status and Future Directions in Boiling Heat Transfer and Two-Phase Flow, p.135 - 141, 2000/00

no abstracts in English

Analyses on pressure wave propagation and dynamic stress in mercury targets

Ishikura, Shuichi*; Kaminaga, Masanori; *; Hino, Ryutaro; *; *; *

Proc. of 14th Meeting of the Int. Collaboration on Advanced Neutron Sources (ICANS-14), 1, p.288 - 300, 1998/00

no abstracts in English

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.

An evaluation of Bernoulli effect on slugging in horizontal two-phase flow

Nakamura, Hideo; Kukita, Yutaka; *

Journal of Nuclear Science and Technology, 31(2), p.113 - 121, 1994/02

https://doi.org/10.3327/jnst.31.113

no abstracts in English

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

Experimental study on DNB heat flux correlations for JMTR safety analysis

; Kaminaga, Masanori; ; Ando, Hiroei; ; Sudo, Yukio;

ANL/RERTR/TM-18, CONF-9009108, 0, p.241 - 248, 1993/00

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