著者

佃 由晃 林 洋 上村 勝一郎 服部 年逸 金子 浩久 師岡 慎一 光武 徹 秋葉 美幸 安部 信明 藁科 正彦 増原 康博 木村 次郎 田辺 朗 西野 祐治 井坂 浩順 鈴木 理一郎

出版者

Atomic Energy Society of Japan

雑誌

日本原子力学会和文論文誌 (ISSN:13472879)

巻号頁・発行日

vol.1, no.4, pp.384-403, 2002

被引用文献数

5

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Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8×8, 9×9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPECTH-B Project). The high-burnup 8×8 fuel (average fuel assembly discharge burnup: about 39.5GWd/t), has been utilized from 1991. And the 9×9 fuel (average fuel assembly discharge burnup: about 45GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9×9 fuel assembly.<BR>Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9×9 fuel combined with the previously reported results of high-burnup 8×8 fuel.<BR>As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed.

著者

山本 泰 武内 豊 白川 健悦 師岡 慎一

出版者

一般社団法人日本機械学会

雑誌

日本機械学會論文集. B編 (ISSN:03875016)

巻号頁・発行日

vol.75, no.751, pp.397-399, 2009-03-25

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In the envisioned high-power density core plant, the degradation of stability is concerned that is attributed to the shortening of rod heat conduction time constant upon the dense fuel grid and the increment of pressure drop caused by the wall friction. Under power and flow oscillating conditions, the two-phase flow behavior might be different from that under the steady conditions. Therefore, the thermal-hydraulic test using the rod bundle under oscillating conditions were performed to obtain the verification data for the analysis code. In this paper, the test results were described. It was found that the critical power decreased under flow oscillating conditions compared with under steady conditions, and the power oscillation had a little effect on the critical power.