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高速炉の炉停止失敗事象における炉容器内終息(IVR)に関する検討,1; ATWSにおけるIVR評価の概要

鈴木 徹; 曽我部 丞司; 飛田 吉春; 堺 公明*; 中井 良大

日本機械学会論文集(インターネット), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

高速炉の炉停止失敗事象(ATWS: Anticipated Transient without Scram)に対して、原子炉容器内での事象終息(IVR: In-Vessel Retention)の成立性を検討した。検討においては、確率論的評価に基づいて冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)をATWSの代表事象に選定した上で、総合的安全解析コードや個別物理モデルを活用して炉心損傷時の事象進展を解析し、事故の機械的影響と熱的影響を評価した。本検討の結果から、原子炉容器は機械的にも熱的にも損傷することはなく、IVRが成立する見通しを得ることができた。


Experimental discussion on fragmentation mechanism of molten oxide discharged into a sodium pool

松場 賢一; 神山 健司; 豊岡 淳一; 飛田 吉春; Zuev, V. A.*; Kolodeshnikov, A. A.*; Vasilyev, Y. S.*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05



Safety requirements expected for the prototype fast breeder reactor "Monju"

齋藤 伸三; 岡本 幸司*; 片岡 勲*; 杉山 憲一郎*; 村松 健*; 一宮 正和*; 近藤 悟; 与能本 泰介

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05

Japan Atomic Energy Agency set up "Advisory Committee on Monju Safety Requirements" in order to establish original safety requirements expected to the prototype FBR "Monju" SFRs use sodium as coolant. It is not necessary to increase primary system pressure for power generation because of the high boiling point of sodium (883$$^{circ}$$C at atmospheric pressure) and sodium coolant liquid level can be maintained by guard vessels. It would therefore not result in core uncovering accident in early stage even in the case of a loss of primary coolant accident which could occur in LWRs, and hence forced pressure reduction and coolant injection are not necessary for SFRs. Liquid sodium can be used in the wide temperature range and be cooled by natural circulation. In addition, multiple accident management strategies by manual operation can be applied because temperature increase is generally gradual even under accident conditions and grace period (several to several ten hours) before significant core damage occurs can be achieved due to large heat capacity of sodium in systems. This paper summarizes the above mentioned safety requirements expected to Monju discussed by the committee.


A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

鈴木 徹; 飛田 吉春; 川田 賢一; 田上 浩孝; 曽我部 丞司; 松場 賢一; 伊藤 啓; 大島 宏之

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 被引用回数:11 パーセンタイル:10.49(Nuclear Science & Technology)

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss-of-flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of In-Vessel Retention (IVR) for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of IVR against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.


Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

丸山 結; 山野 憲洋; 森山 清史; H.S.Park*; 工藤 保; Y.Yang*; 杉本 純

NEA/CSNI/R(98)18, p.243 - 250, 1999/02


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