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論文

Numerical simulation of heat transfer behavior in EAGLE ID1 in-pile test using finite volume particle method

Zhang, T.*; 船越 寛司*; Liu, X.*; Liu, W.*; 守田 幸路*; 神山 健司

Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01

The EAGLE ID1 test was performed by the Japan Atomic Energy Agency to demonstrate the effectiveness of fuel discharge from a fuel subassembly with an inner duct structure. The experimental results suggested that the early duct wall failure observed in the test was initiated by high heat flux from the molten pool comprising liquid fuel and steel. In addition, the post-test analyses showed that the high heat flux may be enhanced effectively by molten steel in the pool. In this study, a series of thermal-hydraulic behaviors in the ID1 test was analyzed to investigate the mechanisms of molten pool-to-duct wall heat transfer using a fully Lagrangian approach based on the finite volume particle method. The present 2D particle-based simulation demonstrated that a large thermal load on the duct wall can be caused by direct contact of the liquid fuel with nuclear heat and high-temperature liquid steel.

論文

Particle-based simulation of heat transfer behavior in EAGLE ID1 in-pile test

守田 幸路*; 小川 竜聖*; 時岡 大海*; Liu, X.*; Liu, W.*; 神山 健司

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10

EAGLE炉内ID1試験は日本原子力研究開発機構によって実施され、FAIDUSと称される内部ダクト付き燃料集合体からの早期燃料流出を模擬したものである。試験で生じた早期ダクト破損は、燃料とスティールから構成される溶融プールからの高い熱流束によるものと解釈されている。試験後の分析からは、壁面に燃料クラストが形成されない状況において、高い熱伝導度を有するプール中の溶融スティールによって溶融プールからダクトへの伝熱が効果的に促進されたことが示唆されている。本研究では、多成分多相流の粒子法に基づいた完全ラグランジェ法を用いて溶融プールからダクト壁への熱伝達機構を分析した。プール中の溶融スティールと燃料の混合と分離挙動およびこれらの挙動がプールからダクトへの伝熱に与える影響を調べるため、燃料ピンの崩壊、溶融プールの形成およびダクト壁の破損に至る一連の挙動を模擬した。現在の2次元粒子法シミュレーションでは、10MW/m$$^{2}$$を超える壁面への大きな熱負荷は、核発熱を伴う液体燃料が壁面へ直接接触することによるものであることが示された。

論文

A Numerical study on local fuel-coolant interactions in a simulated molten fuel pool using the SIMMER-III code

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Annals of Nuclear Energy, 85, p.740 - 752, 2015/11

 被引用回数:19 パーセンタイル:7.56(Nuclear Science & Technology)

Studies on local fuel-coolant interactions (FCI) in a molten pool are crucial to the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs). To clarify the characteristics of this interaction, in recent years a series of simulated experiments, which covers a variety of conditions including much difference in water volume, melt temperature, water subcooling and water release site (pool surface or bottom), was conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, motivated by acquiring further evidence for understanding its mechanisms, interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency, are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is confirmed that, similar to experiments, the water volume, melt temperature and water release site are observable to have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the most probable reason leading to the limited pressurization and resultant mechanical energy release for a given melt and water temperature within the non-film boiling range, even under a condition of much larger volume of water entrapped within the pool, should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

論文

First analysis of local fuel-coolant interactions in a molten pool by SIMMER-III using reactor materials

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

To clarify the mechanisms underlying local fuel-coolant interactions (FCI) in a molten pool, in this study, several latest calculations with reactor materials were performed using SIMMER-III, an advanced fast reactor safety analysis code. The performed SIMMER-III analyses suggest that despite of a comparatively larger temperature range of molten-fuel and sodium possibly varied during reactor accidents, the isolation effect of vapor bubbles generated at the melt-sodium interface seems to be the unique dominant mechanism that leads to the limited pressurization. Knowledge and fundamental data from this work might be utilized for future empirical-approach studies (e.g. those investigating the characteristics of critical coolant volume required for achieving the saturated pressurization at varied melt and coolant temperatures).

論文

SIMMER-III analyses of local fuel-coolant interactions in a simulated molten fuel pool; Effect of coolant quantity

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00

 被引用回数:1 パーセンタイル:86.4(Nuclear Science & Technology)

To clarify the mechanisms underlying local fuel-coolant interactions (FCI) in a molten pool, in recent years several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

報告書

シビアアクシデントの伝熱流動現象における素過程に関する研究; 溶融炉心プールと冷却水との液滴界面における熱伝達, 原子力基礎研究 H10-027-6 (委託研究)

三島 嘉一郎*; 齋藤 泰司*

JAERI-Tech 2002-014, 83 Pages, 2002/03

JAERI-Tech-2002-014.pdf:6.83MB

シビアアクシデント時の溶融燃料プールと冷却水との液液界面における熱伝達の把握を目的として、溶融ウッズメタルと蒸留水とを用いた定常及び非定常熱伝達実験を行った。定常実験では、自然対流領域から膜沸騰領域に至る沸騰曲線を取得するとともに、沸騰挙動を高速度ビデオにより観察した。非定常実験では、高温の溶融金属上に蒸留水を注入し、冷却過程における沸騰曲線を得た。得られた沸騰曲線を、固液系及び液液系に対する既存の相関式や実験データと比較し、以下の結論を得た。(1)界面の揺動が無視でき、かつ、界面に酸化膜に形成される場合には、液液系の沸騰曲線は、固液系の核沸騰及び膜沸騰領域の熱伝達相関式並びに限界熱流束相関式により概ね予測できる。(2)液液界面に酸化物が存在しない場合には、Novakovicらの水銀を用いた実験結果と同様、液液系の沸騰熱伝達は固液系の沸騰曲線により高過熱度側に移行する。(3)非定常状態における膜沸騰において、熱伝達率は、固液系の膜沸騰に対する推算値より約100%程度大きい値を示した。これは、界面全体の激しい揺動のために、みかけの熱伝達率が増大したものと考えられる。

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