著者

白石 直 渡壁 壽人 佐郷 ひろみ 中村 友道 石谷 嘉英 此村 守 山口 彰 藤井 正

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

vol.2004, no.7, pp.103-104, 2004-09-04

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A flow-induced vibration test facility that simulates a hot leg piping of the Japanese sodium-cooled fast reactor (JSFR) with 1/3 scale is used to investigate the flow pattern and pressure fluctuations of the pipe. As the observation of flow pattern, the velocity distributions measured by LDV show the flow pattern is independent of Reynolds number at high Reynolds number. The maximum velocity is about 15 times the mean velocity in the elbow. A statistical analysis of the pressure fluctuations in a separation region shows the power spectrum is of white noise up to 20Hz, the autocorrelation sharply drops to zero less than 1 sec of time interval and the probability density distribution figures almost the Gaussian distribution, excepting its flatness-3 is 3.

著者

佐郷 ひろみ 町田 秀夫 和田 宏 加口 仁 田辺 宏暁

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

no.1, pp.383-384, 2001-08-22

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A part of the SORE system was developed for Reactor Vessel. SORE is prototype system that calculates stress history and creep-fatigue damage for main component of MONJU using plant data In this study, the calculating method of stress and temperature using plant data and the procedure of fatigue damage and creep damage evaluation using calculated history of temperature and stress was investigated.

著者

田中 正暁 佐郷 ひろみ 岩本 幸治 江原 真司 小野 綾子 村上 貴裕 早川 教

出版者

The Japan Society of Mechanical Engineers

雑誌

日本機械学会論文集 B編 (ISSN:03875016)

巻号頁・発行日

vol.78, no.792, pp.1392-1396, 2012

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A study on flow induced vibration in the primary cooling system of Japan Sodium cooled Fast Reactor (JSFR) consisting of large diameter pipe and pipe elbow with short curvature radius ("short-elbow") has been conducted. Flow-induced vibration in the short-elbow is an important issue in design study of JSFR, because it may affect to structural integrity of the pipe. In this paper, unsteady flow characteristics in the JSFR short-elbow pipe related to the large-scale eddy motion were estimated based on knowledge from existing studies for curved pipes and scaled water experiments and numerical simulations for the JSFR hot-leg piping.

著者

藤井 正 西口 洋平 此村 守 佐郷 ひろみ 白石 直 渡壁 壽人 中村 友道 石谷 嘉英

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

vol.2004, no.3, pp.247-248, 2004-09-04

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A conceptual design study of the sodium-cooled fast reactor (JSFR) is in progress in "Feasibility Study on Commercialized Fast Reactor Cycle Systems". The cooling system of the reactor is composed of two loops in order to reduce plant construction cost. According to reduction of loop number, large diameter pipings are adopted in the primary cooling system, and the average sodium velocity in the piping increases to 9 m/s level. One of issues for piping design is to confirm hydraulic and flow-induced vibration behaviors of the piping under high Reynolds number (10^7 order level) conditions. Then, a flow-induced vibration test facility which simulates a hot leg piping with 1/3 scale has been fabricated. As a first step of the test series, this report describes outline of flow visualization test results.

著者

石谷 嘉英 中村 友道 佐郷 ひろみ 白石 直 渡壁 壽人 此村 守 山口 彰 藤井 正

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

vol.2004, no.7, pp.105-106, 2004-09-04

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A 1/3scale flow-induced vibration test facility that simulates a hot-leg piping of the Japanese sodium-cooled fast reactor (JSFR) is used to investigate the pressure fluctuations of the pipe. To evaluate the flow-induced vibrations for the hot-leg and cold-leg pipes, the random force distributions along the pipe and their correlations are estimated. As the result, the power spectrum densities of pressure fluctuations are classified into four sections, the correlation lengths of axial direction into three sections, and the correlation lengths of tangential direction into four sections. The maximum flow-induced random vibration force in the pipe is estimated in the region of flow separation downstream the elbow.

著者

田辺 宏暁 渡士 克己 平山 浩 佐郷 ひろみ 芋生 和道

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

no.1, pp.381-382, 2001-08-22

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Plant maintenance and plant management activities are important for steady and safety operation of the nuclear plant. Therefore the Structural Integrity Oriented Reliability Assessment System (SORE) has developed to assist preservation management for the main components of MONJU plant. SORE calculates stress intensity and creep-fatigue damage of main parts of components using the plant data (temperature history, etc.).

著者

中村 友道 佐郷 ひろみ 白石 直 此村 守 山口 彰 藤井 正

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

vol.2004, no.7, pp.101-102, 2004-09-04

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1/10^<th> scale model flow test has been conducted to measure the fluid force acting on the upper internal structure in a sodium reactor. This model has already been constructed and it couldn't be re-arranged including its flow section. Then, the internal structure is supported with flexible rods to measure the response, preventing the effect on the flow pattern. The exciting force by flowing fluid is analyzed by solving an inverse problem, where the exciting force is obtained by screening the vibration characteristics from the responding vibration signal. Because the real exciting force has not known, two methods, from the acceleration response and from the strain gage attached on the supporting rods, are tried to obtain the same force, and they show a reasonable coincidence