Influence of steam flow rate on oxidation kinetics of silicon carbide at 1400-1600 C

Pham, V. H.; Kurata, Masaki; Nagae, Yuji; Ishibashi, Ryo*; Sasaki, Masana*

Corrosion Science, 255, p.113098_1 - 113098_9, 2025/10

https://doi.org/10.1016/j.corsci.2025.113098

Achievement of safety demonstration tests using HTTR; Loss of forced cooling test at 100% reactor power (30 MW)

Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.

JAEA-Research 2025-005, 23 Pages, 2025/07

JAEA-Research-2025-005.pdf:2.68MB

A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.

Comparison of two-phase critical flow models for estimation of leak flow rate through cracks

Watanabe, Tadashi*; Katsuyama, Jinya; Mano, Akihiro

International Journal of Nuclear and Quantum Engineering (Internet), 13(11), p.516 - 519, 2019/10

The estimation of leak flow rates through narrow cracks in structures is of importance for nuclear reactor safety, since the leak flow could be detected before occurrence of loss-of-coolant accidents. The two-phase critical leak flow rates are calculated using the system analysis code, and two representative non-homogeneous critical flow models, Henry-Fauske model and Ransom-Trapp model, are compared. The pressure decrease and vapor generation in the crack, and the leak flow rates are found to be larger for the Henry-Fauske model. It is shown that the leak flow rates are not affected by the structural temperature, but affected largely by the roughness of crack surface.

Research and development on lead-bismuth technology for accelerator-driven transmutation system at JAERI

Kurata, Yuji; Kikuchi, Kenji; Saito, Shigeru; Kamata, Kinya*; Kitano, Teruaki*; Oigawa, Hiroyuki

Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.267 - 277, 2004/05

R&D on lead-bismuth technology have been conducted for accelerator driven system. From the test results of 316SS using JLBL-1 for 3000h without active oxygen control, mass transfer from high temperature to low temperature parts was observed. It was found that deposition of Pb/Bi and Fe-Cr grains in the annular channel of the electro-magnetic pump caused plugging and decrease in flow rate. The modification of the loop system brought about a good effect on operation. Significant erosion/corrosion was not observed in the experiment using MES loop for 1000h under 10wt.% oxygen condition. The results of static corrosion tests showed the following: corrosion depth decreased at 450C with increasing Cr content in steels while corrosion depth of JPCA and 316SS became larger due to ferritization caused by dissolution of Ni and Cr at 550C. Si-added steel exhibited good corrosion resistance at 550C.

Research on the fundamental process of thermal-hydraulic behaviors in severe accident; Boiling heat transfer in a narrow channel, JAERI's nuclear research promotion program, H10-027-4 (Contract research)

Kaminaga, Fumito*

JAERI-Tech 2002-012, 68 Pages, 2002/03

JAERI-Tech-2002-012.pdf:2.45MB

no abstracts in English

Supercritical helium pump

Kato, Takashi

Tabo Kikai, 28(9), p.536 - 545, 2000/09

https://doi.org/10.11458/tsj1973.28.536

no abstracts in English

The Review of the application of neutron radiography to thermal hydraulic research

Mishima, Kaichiro*; Hibiki, Takashi*; Saito, Yasushi*; Nakamura, Hideo; Matsubayashi, Masahito

Nuclear Instruments and Methods in Physics Research A, 424(1), p.66 - 72, 1999/00

https://doi.org/10.1016/S0168-9002(98)01240-6

no abstracts in English

High-frame rate neutron radioscopy with a steady thermal neutron beam

Mishima, Kaichiro*; Hibiki, Takashi*; Fujine, Shigenori*; Yoneda, Kenji*; Kanda, Keiji*; Nishihara, Hideaki*; Matsubayashi, Masahito; Tsuruno, Akira

Fifth World Conf. on Neutron Radiography, 0, p.140 - 147, 1996/00

no abstracts in English

Measurement of local void fraction distribution in rod bundle under high-pressure high-temperature boil-off conditions by using optical void probe

Kumamaru, Hiroshige; Murata, Hideo; ; Kukita, Yutaka

The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.217 - 222, 1995/00

no abstracts in English

Visualization of fluid phenomena using a high frame-rate neutron radiography with a steady thermal neutron beam

Hibiki, Takashi*; Mishima, Kaichiro*; Yoneda, Kenji*; Fujine, Shigenori*; Tsuruno, Akira; Matsubayashi, Masahito

Nuclear Instruments and Methods in Physics Research A, 351, p.423 - 436, 1994/00

https://doi.org/10.1016/0168-9002(94)91371-4

no abstracts in English