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安部 諭; 柴本 泰照
Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high temperature flow of approximately 390
C was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high temperature region. The phenomenological discussion in this paper help understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.
下村 健太; 山下 拓哉; 永江 勇二
JAEA-Data/Code 2022-012, 270 Pages, 2023/03
JAEA-Data-Code-2022-012.pdf:38.25MB
発電用原子炉である軽水炉において、東京電力ホールディングス株式会社福島第一原子力発電所と同様な全交流電源喪失が発生した場合には、原子炉圧力容器(RPV: Reactor Pressure Vessel)内の冷却機能の喪失、炉内の水位低下による燃料棒の露出、炉心溶融に伴うRPVの破損やRPV破損に伴う炉内の放射線物質の漏えいが発生することが考えられる。事故進展におけるRPVの損傷、溶融した燃料デブリの流出・拡大等の過程を検証、推定することは、廃炉作業を進める上で重要な情報となる。RPVの破損要因については、RPV下部構造部に加えられる荷重・拘束に起因する破損(力学的破損)、低融点金属や高融点酸化物とRPV底部の構造部材との共晶現象による破損(材料間反応による破損)、RPV底部の構造部材の融点近傍での破損が考えられる。破損要因の内、力学的破損については、数値解析(熱流動解析及び構造解析)により検証を行う。このような数値解析を実施する際には、RPV及び炉内構造物を構成する材料(ジルコニウム,炭化ホウ素,ステンレス鋼,ニッケル合金,低合金鋼等)の融点近傍までの伝熱特性(熱伝導率,比熱,密度)や材料特性(熱膨張係数,ヤング率,ポアソン比,引張,クリープ)が必要となる。本資料においては、公開文献を基に、RPV及び炉内構造物を構成する各材料の融点近傍までの母材の特性データをデータ集として取りまとめた。なお、RPV及び炉内構造物を構成する構造物の中には溶接部も存在するため、今回限られたデータであるが、溶接部に関する特性データも併せて示した。
Zhang, T.*; 守田 幸路*; Liu, X.*; Liu, W.*; 神山 健司
Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12
被引用回数：1 パーセンタイル：31.61(Nuclear Science & Technology)The ID1 test was the final target test of the EAGLE experimental framework program. It was used to verify that during a core disruptive accident, the molten fuel could be discharged via wall failure of an inner duct in FAIDUS, a design concept for the sodium-cooled fast reactor. The ID1 results revealed that the wall failure behavior owed to the large heat flow from the surrounding fuel/steel mixture. The present study numerically investigated the heat transfer mechanisms in the test using the finite volume particle method in the three-dimensional domain. The thermal hydraulic behaviors during wall failure were reproduced reasonably. The present three-dimensional simulation mitigated inherent defects of our previous two-dimensional calculation and clarified that the solid fuel and liquid steel close to the outer surface of the duct can expose the duct to high thermal loads, resulting in the wall failure.
安部 諭; Hamdani, A.; 石垣 将宏*; 柴本 泰照
Annals of Nuclear Energy, 166, p.108791_1 - 108791_18, 2022/02
被引用回数：5 パーセンタイル：56.94(Nuclear Science & Technology)This paper describes an experimental investigation of natural convection driven by outer surface cooling in the presence of density stratification consisting of an air-helium gas mixture (as mimic gas of hydrogen) in an enclosed vessel. The unique cooling system of the Containment InteGral effects Measurement Apparatus (whose test vessel is a cylinder with 2.5-m diameter and 11-m height) is used, and findings reveal that the cooling location relative to the stratification plays an important role in determining the interaction behavior of the heat and mass transfer in the enclosed vessel. When the cooling region is narrower than the stratification thickness, the density-stratified region expands to the lower part while decreasing in concentration (stratification dissolution). When the cooling region is wider than the stratification thickness, the stratification is gradually eroded from the bottom with decreasing layer thickness (stratification breakup). This knowledge is useful for understanding the interaction behavior of heat and mass transfer during severe accidents in nuclear power plants.
Zhang, T.*; 守田 幸路*; Liu, X.*; Liu, W.*; 神山 健司
Extended abstracts of the 2nd Asian Conference on Thermal Sciences (Internet), 2 Pages, 2021/10
日本のナトリウム冷却高速炉では、高速炉の炉心損傷事故における大規模炉心プール形成による再臨界を回避する方策として、内部ダクト付き燃料集合体(FAIDUS)が提案されている。本研究では、FAIDUSの有効性を実証するために実施されたEAGLE ID1炉内試験を対象に3次元粒子粒子法シミュレーションを行い、溶融燃料/スティールの混合プールからダクト壁への熱伝達機構を明らかにするための解析的検討を行った。
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
被引用回数：5 パーセンタイル：65.59(Nuclear Science & Technology)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.
Wang, Z.; Duan, G.*; 松永 拓也*; 杉山 智之
International Journal of Heat and Mass Transfer, 157, p.119919_1 - 119919_20, 2020/08
被引用回数：16 パーセンタイル：77.06(Thermodynamics)Enforcing accurate and consistent boundary conditions is a difficult issue for particle methods, due to the lack of information outside boundaries. Recently, consistent Neumann boundary condition enforcement is developed for the least squares moving particle semi-implicit method (LSMPS). However, the Robin boundary cannot be straightforwardly considered by that method because no computational variables are defined on the wall boundary. In this paper, a consistent Robin boundary enforcement for heat transfer problem is proposed. Based on the Taylor series expansion, the Robin boundary condition for temperature is converted to the fitting function of internal rather than boundary particles and incorporated into least squares approach for discretization schemes. Arbitrary geometries can be easily treated due to the use of polygons for wall boundary. A convergence study was firstly carried out to verify the consistency. Then, numerical tests of 1-D and 2-D heat conduction problems subjected to mixed boundary conditions were performed for verification, and good agreements with theoretical solutions were observed. Natural convection problems with different boundary conditions in an annulus were carried out for further validations of heat-fluid coupling. Excellent agreements between the present and literature results were demonstrated.
阿部 雄太; 山下 拓哉; 佐藤 一憲; 中桐 俊男; 石見 明洋
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04
The authors are developing an experimental technology for simulating severe accident (SA) conditions using simulate fuel material (ZrO
) that would contribute, not only to Fukushima Daiichi (1F) decommissioning, but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of accident progression behavior. Nontransfer (NTR) type plasma, which has been in practical use with a large torch capacity as high as 2 MW, has the potential to heat subject materials to very high temperatures without selecting the target to be heated. When simulating 1F with SA code, the target of this core-material-melting and relocation (CMMR) experiment was to confirm that NTR plasma has a sufficient heating performance realizing large temperature gradients (
2000 K/m) expected under 1F conditions. The authors selected NTR-type plasma-heating technology that has the advantage of continuous heating in addition to its high-temperature level. The CMMR-2 experiments were carried out in 2017 applying the improved technology (higher heating power and controlled oxygen concentration). The CMMR-2 experiment adopted a 30-min heating period, wherein the power was increased to a level where a large temperature gradient was expected at the lower part of the core under actual 1F accident conditions. Most of the control blade and channel box migrated from the original position. After heating, the simulated fuel assembly was measured by X-ray computed tomography (CT) technology and by electron probe micro-analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective, in terms of applicability of the NTR-type plasma-heating technology to the SA experimental study, was obtained.
小川 益郎*
JAEA-Technology 2019-010, 22 Pages, 2019/07
JAEA-Technology-2019-010.pdf:1.5MB
円管内流れは、流れが実際に遷移し、遷移流が間欠性を示すにもかかわらず、あらゆる小さな外乱に対して線形的に安定である。このことは、流体力学ではまだ解決されていない大きな課題の一つである。そこで、著者は、これまで誰も気がつかず認識してこなかった事実を初めて指摘する。この事実というのは、「円管内の流れは、流れの剥離によって、円管入り口付近に形成される剥離泡から放出された渦のために層流から遷移し、そして渦放出が間欠的であるために遷移流が間欠性を示す。」というものである。この事実は、円管の入口形状が遷移レイノルズ数に大きく影響することや、第3の遷移現象に分類されている外側円筒が支配的に回転する同心二重円筒間の流れが円管内の遷移流れと同様に流れの剥離によって間欠性を示すといった、多くの実験結果によって裏付けられている。本研究によって、高温ガス冷却炉の熱流体設計において最も重要な課題の一つである熱伝達促進のために、急縮小型の入口形状が遷移開始レイノルズ数をできる限り小さくできることを明らかにした。
上澤 伸一郎; 山下 晋; 柴田 光彦; 吉田 啓之
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
A dry method for fuel debris is proposed for decommissioning of TEPCO's Fukushima Daiichi NPS. We have been evaluating the air-cooling performance of the fuel debris in the dry method by using JUPITER. Because JUPITER can calculate relocation of the corium, it is expected to calculate thermal-hydraulic simulation of the air cooling of the fuel debris in the dry method based on the calculated debris position, shapes and composition with the relocation analysis. In this paper, the experiment of heat transfer and flow visualization of free convection adjacent to upward-facing horizontal heat transfer surface was performed to validate the calculation of the free convective heat transfer with JUPITER. In the experiment, the temperature distribution was measured with a thermocouple tree. In addition, the velocity distribution of free convection was visualized by a particle image velocimetry (PIV). In the comparison between the JUPITER and the experiment, the temperature distribution for the vertical direction in the quasi-steady state was fitted between the JUPITER and the experiment. The velocity distribution calculated with JUPITER was also in good agreement with the experimental result. Therefore, it is expected that JUPITER is a helpful numerical method to evaluate the air-cooling performance of the fuel debris in the dry method.
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
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。
江口 悠太; 菅原 隆徳; 西原 健司; 田澤 勇次郎; 辻本 和文
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
加速器駆動核変換システム(ADS)の基礎核特性研究のため、J-PARC計画において核変換物理実験施設(TEF-P)の建設が検討されている。本研究では、崩壊熱の大きなマイナーアクチノイド(MA)燃料を多く使用するTEF-Pにおいて、炉心冷却システムが停止した場合の自然冷却特性の評価、及びその際に炉心が損傷しない設計条件検討を行った。TEF-Pの炉心温度評価においては、炉心周辺部の空格子管領域が断熱層として大きく影響を及ぼすことから、空格子管領域の熱伝達特性を測定するモックアップ試験装置を製作して実験を行い、実験的な熱伝達率を得た。この結果を元に、TEF-P炉心の三次元伝熱解析を実施し、制限温度である327Cを下回る294Cという結果を得た。
阿部 雄太; 山下 拓哉; 佐藤 一憲; 中桐 俊男; 石見 明洋; 永江 勇二
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
Authors are developing an experimental technology to realize experiments simulating Severe Accident (SA) conditions using simulant fuel material (ZrO with slight addition of MgO for stabilization) that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. Based on the results of the prototype test, improvement of plasma heating technology was conducted. The Core Material Melting and Relocation (CMMR)-1/-2 experiments were carried out in 2017 with the large-scale simulated fuel assembly (1 m 
0.3 m 
) applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different resulting basically in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment is selected here from the viewpoint of establishing an experimental technology. The CMMR-2 experiment adopted 30-min heating period, the power was increased up to a level so that a large temperature gradient ( 2,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. Most of the control blade and the channel box migrated from the original position. After the heating, the simulated fuel assembly was measured by the X-ray Computed Tomography (CT) technology and by Electron Probe Micro Analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective in terms of applicability of the non-transfer type plasma heating technology to the SA experimental study was obtained.
丸山 信一郎*; 綿谷 聡*
三井住友建設技術研究開発報告, (15), p.107 - 112, 2017/10
福島第一原子力発電所(以下、1Fと称す)の廃止措置において、安全で確実な燃料デブリの取出しを行うためには、燃料デブリの形態や特性を推定することが不可欠となる。その推定のため、事故時の燃料集合体の溶融移行挙動調査が行われている。調査にあたり、燃料集合体溶融模擬材の切断が必要となり、切断にはジルコニウム合金とステンレスの溶融混合材料やセラミックの切断実績のあるアブレイシブウォータージェット(以下、AWJと称す)工法を適用することとした。結果、燃料集合体溶融模擬材を切断でき、切断可能な条件のデータを取得できた。今後、そのデータは、燃料デブリの取出しの検討に役立てることができる。
阿部 雄太; 佐藤 一憲; 中桐 俊男; 石見 明洋; 永江 勇二
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Authors are developing an experimental technology to realize experiments simulating severe accident conditions that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. In the first part of this program, called Phase I hereafter, a series of small-scale experiments (10 cm 10 cm 25 cmh) were performed in March 2015 and it was demonstrated that non-transfer (NTR) type plasma heating is capable of successfully melting the high melting-point ceramics. In order to confirm applicability of this heating technology to larger scale test specimens to address the experimental needs, authors performed a second series plasma heating tests in 2016, called Phase II hereafter, using a simulated fuel assembly with a larger size (100 cm 30 cm phi). In the phase II part of the program, the power was increased up to a level so that a large temperature gradient (2,000 K/m - 4,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. After the heating, these test pieces were measured by the X-ray Computed Tomography (CT) technology. CT pictures demonstrated its excellent performance with rather good precision. Based on these results, basic applicability of the NTR plasma heating for the SA experimental study was confirmed. With the Phase II-type 100 cm-high test geometry, core material relocation (CMR) behavior within the active core region and its access to the core support structure region would be addressed. JAEA is also preparing for the next step large-scale tests using up to four simulated fuel assemblies covering the lower part of the active fuel and fully simulating the upper part of the lower core support structures addressing CMR behavior including core material relocation into the lower plenum.
秋本 肇; 菅原 隆徳
JAEA-Data/Code 2016-008, 87 Pages, 2016/09
JAEA-Data-Code-2016-008.pdf:15.62MB
鉛ビスマス冷却加速器駆動システム(ADS)の基本設計に資するため、定格運転時の熱流動解析を行った。概念設計で得られた機器の性能諸元と寸法を整理し、ADS設計解析コード用入力データを作成した。急峻な半径方向出力分布を有するADSの炉心部分を詳細にモデル化し、炉心内の3次元的な流体混合が炉心冷却に与える影響を評価した。定格運転時の熱流動解析の結果から、(1)定格運転時の最高被覆管表面温度と最高燃料中心温度は設計制限値を下回る。(2)燃料集合体間の冷却材流量配分に対する半径方向出力分布の影響は小さい。急峻な半径方向出力分布がある場合でも炉心加熱区間入口における冷却材流量分布はほぼ平坦である。(3)燃料棒表面における熱伝達率に対する半径方向出力分布の影響は小さい。出力の違いに伴う被覆管表面温度の差異は、主に燃料棒に隣接する冷却材温度の違いにより決定される。(4)蒸気発生器4基における熱水力学的挙動は対称である。また、主循環ポンプ2基における熱水力学的挙動も対称である。ことがわかった。詳細な計算で明らかとなって熱水力学的挙動を踏まえて入力データを簡素化した簡易モデルを作成した。
阿部 雄太; 佐藤 一憲; 石見 明洋; 中桐 俊男; 永江 勇二
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
原子力機構では非移行型プラズマ加熱を用いたBWR体系での炉心物質の下部プレナムへの移行挙動(CMR)に着目した試験を検討している。この技術の適用性を確認するため、我々は小規模試験体(107mm107mm222mmh)を用いたプラズマ加熱の予備実験を行った。これらの予備実験の結果から、SA(シビアアクシデント)研究への非移行型プラズマ加熱の優れた適用可能性が確認できた。また我々は、2016年に中規模の予備実験(燃料ピン50ロッド規模)を準備し、まだ技術的な適用性が確認できていない制御ブレードやCMR自体に関する試験を実施予定である。
上澤 伸一郎; 小泉 安郎; 吉田 啓之
Proceedings of 9th International Conference on Multiphase Flow (ICMF 2016) (CD-ROM), 6 Pages, 2016/05
Since seawater was injected into cores in the accident at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company, pool nucleate boiling heat transfer experiments of the distilled water, the real seawater and the 3.5-10 wt% artificial seawater were conducted to examine the effect of sea-salt-deposition on a heat transfer surface. In the seawater experiments, the sea salt which was calcium sulfate was deposited on the heat transfer surface at a certain heat flux. At the same time, the temperature of the heat transfer surface kept rising. The surface temperature rose above 473 K although input heating power was constant. The heat transfer surface temperature rising was caused by the growth of the deposited salt layer, because the deposited calcium sulfate has quite low thermal conductivity. In addition, the relation between the concentration of seawater and the heat flux when the calcium sulfate grew on the heat transfer surface indicated that the deposition occurred by the vaporization in the vicinity of the heat transfer surface.
江口 悠太; 菅原 隆徳; 西原 健司; 田澤 勇次郎; 井上 昭; 辻本 和文
JAEA-Technology 2015-052, 34 Pages, 2016/03
JAEA-Technology-2015-052.pdf:5.02MB
大強度陽子加速器施設J-PARC計画で建設が予定されている核変換物理実験施設(TEF-P: Transmutation Physics Experimental Facility)では、MAを含む崩壊熱の大きな燃料を大量に取り扱うため、炉心冷却ブロワ停止時の炉心温度上昇の評価が不可欠である。冷却ブロワ停止時の炉心温度変化は全炉心熱伝導解析によって評価されるが、燃料及びブランケット領域の外側の空格子管領域に適用されている熱通過パラメータの適用性の検証が必要である。本報告書では、空格子管内部における伝熱量の算出モデルである、鉛直等温平面の上下を断熱材で囲った密閉流体層の自由対流モデルの精度検証のため、TEF-P炉心の格子管形状を模擬した試験装置を製作し、試験装置の格子管内部を通過する一次元熱流と温度分布を測定した。これにより、実際の空格子管内部の等価熱伝導率は密閉流体層の自由対流モデルよりも大きいことが明らかになった。また、空格子管内部にアルミニウムブロックを充填することで空格子管体系よりも高い等価熱伝導率となることを確認した。本実験により、TEF-P炉心の温度評価に供することができる熱通過パラメータを実験的に取得した。
上澤 伸一郎; 永武 拓; Jiao, L.; Liu, W.; 高瀬 和之; 吉田 啓之
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11
To understand the current status of the TEPCO's Fukushima Daiichi Nuclear Power Station, the progress of the accident has been calculated by severe accident analysis codes, for example, MAAP, SAMPSON and so on. However, effects of seawater are not considered in these calculations, although the seawater was attempted to inject into the reactors to cool down the nuclear fuels. In the present study, the objective is to understand the basic physical effect of the seawater on the thermal-hydraulic behavior without boiling. We measured and compared the thermal-hydraulic behavior in pure water, NaCl solution and artificial seawater with the concentration of 3.5wt% in a heat transfer and flow visualization experiment by using an internally heated annulus. Above Re = 2300 [-], the correlations between Nusselt number and Reynolds number in the NaCl solution and the artificial seawater were the same with that in the pure water. Moreover, the correlation can be predicted by Dittus-Boelter equation. Below Re = 2300 [-], the Nusselt numbers of each fluid correlated with the Rayleigh number. Therefore, considering physical properties of the NaCl solution and the artificial seawater, the thermal-hydraulics behavior without boiling in the NaCl solution and the artificial seawater was not different from the behavior in the pure water.
