Experimental study on condensation of steam/air mixture in a horizontal tube

Experimental investigation on natural convection heat transfer characteristics of C-shape heating rods bundle used in PRHR HX

Numerical and experimental investigation on the transient heat transfer characteristics of C-shape rod bundles used in Passive Residual Heat Removal Heat Exchangers

Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, univers...

Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident

Modelling guidelines for core exit temperature simulations with system codes

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

Measurement of non-condensable gas in a PWR small-break LOCA simulation test with LSTF for OECD/NEA ROSA Project and RELAP5 post-test analysis

A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW) transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW) was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG) secondary-side pressures were maintained, respectively, at around 16 and 8 MPa by cycle opening of pressurizer (PZR) power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

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Takeshi Takeda

Papers

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

Measurement of non-condensable gas in a PWR small-break LOCA simulation test with LSTF for OECD/NEA ROSA Project and RELAP5 post-test analysis

RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS