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岡垣 百合亜; 柴本 泰照; 和田 裕貴; 安部 諭; 日引 俊詞*
Journal of Nuclear Science and Technology, 60(8), p.955 - 968, 2023/08
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)Pool scrubbing is an important filtering process that prevents radioactive aerosols from entering the environment in the event of severe accidents in a nuclear reactor. In this process of transporting aerosol particles using bubbles, bubble hydrodynamics plays a crucial role in modeling pool scrubbing and significantly affects particle removal in a bubble. The pool scrubbing code based on Lumped Parameter (LP) approach includes the particle removal model, and its hydrodynamic parameters are determined based on simple assumptions. We aim to apply the three-dimensional Computer Fluid Dynamics (CFD) approach to understand the detailed bubble interaction. This study validated the applicability of the CFD simulation to bubble hydrodynamics at the flow transition from a globule to a swarm region, which is critical in the stand-alone pool scrubbing code-SPARC-90. Two types of solvers based on the Volume Of Fluid (VOF) and the Simple Coupled Volume Of Fluid with Level Set (S-CLSVOF) methods were used to capture the gas-liquid interface in the CFD simulation. We used the experimental data for validation. As a result, the VOF and S-CLSVOF methods accurately predicted the bubble size and void fraction distributions. In addition, we confirmed that the bubble rise velocity of the S-CLSVOF method almost agreed with the experimental results.
安部 諭; 岡垣 百合亜
Nuclear Engineering and Design, 404, p.112165_1 - 112165_14, 2023/04
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)Pressurized Thermal Shock (PTS) is induced potentially by the rapid cooling of the cold-leg and downcomer wall in the primary system of a Pressurized Water Reactor (PWR) due to the initiation of Emergency Core Cooling System (ECCS). Thus, fluids mixing in a horizontal cold-leg and downcomer should be predicted accurately; however, turbulence production and damping often hinders this prediction due to the presence of the density gradients. Hence, the Fifth International Benchmark Exercise, the cold-leg mixing Computational Fluid Dynamics (CFD) Benchmark, was conducted under the support of OECD/NEA. The experiment was designed for visualization of the mixing phenomena of two liquids with different densities. The heavy liquid was a simulant of cold water from ECCS, in a horizontal leg and downcomer. We used the Large-eddy Simulation (LES) to investigate the time fluctuation behaviors of velocity and liquid concentration. The CFD simulation was performed with two turbulence models and three different numerical meshes. We investigated the characteristics of the appearance frequency of the heavy liquid concentration with the statistical method. Based on our findings, we propose further experiments and numerical investigations to understand the fluid mixing phenomena related to PTS.
Dehbi, A.*; Cheng, X.*; Liao, Y.*; 岡垣 百合亜; Pellegrini, M.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 15 Pages, 2022/03
Nuclear degraded cores produce fission product aerosols that may reach the environment if not removed by natural processes and/or filtering equipment. The transport paths of aerosols usually include transits through stagnant water pools. It is therefore essential to develop computational tools to predict the aerosols retention by water pools. Currently, this is mostly done with 1-D lumped-parameter codes that are too simplistic to capture the physics. It is hence worthwhile to attempt the CFD approach, which has recently become reasonably mature to address bubble hydrodynamics in low momentum two-phase flows. In this first comparative exercise, we restrict the investigation to a hypothetical parallelepiped water pool (2
28 cm
) into which air is injected through a circular 4 mm ID orifice at low velocity of 0.2 m/s. We present predictions of the gas phase dynamics (void and velocity profiles) for both Euler-Euler and Interface Tracking (IT, Volume-of-Fluid (VOF)) methodologies. In addition, we compare bubble shape, volume and detachment frequency from various IT simulation codes (CFX, Fluent, Star-CCM+, OpenFOAM). Reasonable agreement is found between IT simulations near the injector, but discrepancies increase as one moves towards the free surface. The disagreement between the Euler-Euler and IT results is substantial throughout the domain. Future studies will consist of validation exercises against experimental data to highlight potential model deficiencies and point to ways of remedying them.
安部 諭; 小尾 善男*; 佐藤 聡; 岡垣 百合亜; 柴本 泰照
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
A modeling of heat transfers with boiling transition (BT) and that after the occurrence of BT, called post-BT, is one of the key technical issues to estimate the duration of surface dryout and the peak cladding temperature during DBA (Design Basis Accident) and BDBA (Beyond Design Basis Accident) in light water reactors. Recently, CFD (Computational Fluid Dynamics) has emerged as a powerful tool for representing the heat transfer mechanism. Our main purpose is to obtain in-depth physical insight into the BT and post-BT phenomena by combining experiment and CFD simulation research. This paper introduces our developing activity for an integrated three-field two-fluid CFD methodology based on the Eulerian-Eulerian approach toward the accurate prediction of the dryout behavior from annular-mist to mist flow regimes. We implemented following interaction terms and functions into OpenFOAM ver.7, an opensource code developed by OpenFOAM foundation, as (1) Interaction terms between liquid film and droplets due to deposition and entrainment, (2) Interaction terms on the liquid interface between the liquid film and the gas phase on friction and heat conduction, (3) Heat transfer from the heated wall to the liquid film, (4) Dryout occurrence judgement and the switching function on the boundary condition. The dryout occurrence judgment is based on a correlation on critical film thickness, which is originally applied into the MARS (Multi-dimensional Analysis of Reactor Safety) code. A trial calculation with the developed solver called two Phase Three Field Euler Foam was performed to check the solver operation. The CFD could simulate temperature increase behavior due to the dryout occurrence, whereas here were still challenges in reproducing the transition from mist flow to single-phase vapor flow.
安部 諭; 岡垣 百合亜; 佐藤 聡; 柴本 泰照
Annals of Nuclear Energy, 159, p.108321_1 - 108321_12, 2021/09
被引用回数：3 パーセンタイル：45.99(Nuclear Science & Technology)Heat transfer in the rod bundle is augmented by the mixing vanes on the spacer grid. We conducted a computational fluid dynamics (CFD) simulation with three isotropic turbulence models - standard 
-epsilon, realizable -epsilon, and SST -omega models - to investigate the relationship between heat transfer and turbulence behaviors downstream a simulant spacer (with four vanes) in a single tube under single-phase flow conditions. Quantitatively, the predicted heat transfer coefficient (HTC), secondary flow intensity, and turbulence intensity with the SST -omega model displayed a better agreement (than the other isotropic models) with the correlation based on previous studies. Furthermore, the turbulence production was localized in the near-spacer region (z/D 
10, where D is the inner diameter), which corresponds with the HTC argumentation region. These results indicate that examining the turbulence production when discussing the HTC augmentation downstream the spacer is essential.
岡垣 百合亜; 与能本 泰介; 石垣 将宏; 廣瀬 意育
Fluids (Internet), 6(2), p.80_1 - 80_17, 2021/02
Many thermohydraulic issues about the safety of light water reactors are related to complicated two-phase flow phenomena. In these phenomena, computational fluid dynamics (CFD) analysis using the volume of fluid (VOF) method causes numerical diffusion generated by the first-order upwind scheme used in the convection term of the volume fraction equation. Thus, in this study, we focused on an interface compression (IC) method for such a VOF approach; this technique prevents numerical diffusion issues and maintains boundedness and conservation with negative diffusion. First, on a sufficiently high mesh resolution and without the IC method, the validation process was considered by comparing the amplitude growth of the interfacial wave between a two-dimensional gas sheet and a quiescent liquid using the linear theory. The disturbance growth rates were consistent with the linear theory, and the validation process was considered appropriate. Then, this validation process confirmed the effects of the IC method on numerical diffusion, and we derived the optimum value of the IC coefficient, which is the parameter that controls the numerical diffusion.
安部 諭; 岡垣 百合亜; 石垣 将宏; 柴本 泰照
Proceedings of OECD/NEA Workshop on Virtual CFD4NRS-8; Computational Fluid Dynamics for Nuclear Reactor Safety (Internet), 11 Pages, 2020/11
The fifth international benchmark exercise (IBE-5), the cold-leg mixing CFD benchmark, was conducted under the support of OECD/NEA. The experiment for IBE-5 was designed to visualize the mixing phenomena of two liquids with different density in a horizontal leg (as a simulant of the cold-leg) and downcomer. This paper shows our CFD result on the open test condition in IBE-5. We selected the Large-eddy simulation (LES) solving the filtered equation of flow and concentration fields. Regarding the eddy-viscosity to model the turbulence flux of the momentum at sub-grid scale (SGS), Wall-adapting locale eddy-viscosity (WALE) model, a modified version from the Smagorinsky model, was applied. The experimental geometry was resolved with three different numerical mesh systems. The CFD analysis predicted the laminar-like flow behavior in the horizontal leg. Due to the large density difference between the two liquids, the turbulence production was suppressed strongly, and the velocity fluctuation in the horizontal leg became very slow and small. In contrast, the strong turbulence mixing in the downcomer was predicted. The plume from the horizontal leg entrained with the surroundings and spread circumferentially in the downcomer. The comparison with the TAMU experimental data reveals the good performance of the WALE model. In addition, we discuss the appearance characteristics of the high concentration of the heavy liquid in the downcomer in the LES. The Probability Density Function (PDF) and Cumulative Distribution Function (CDF) are derived based on the predicted time-series of the heavy liquid concentration. The PDF around the mean concentration in the case with the low mesh resolution is larger than that predicted by the higher resolution due to the excessive homogenization of the heavy fluid concentration. This study reveals the importance to note the required mesh resolution to predict the appearance event of the high concentration.
岡垣 百合亜; 柴本 泰照; 安部 諭
Proceedings of OECD/NEA Workshop on Virtual CFD4NRS-8; Computational Fluid Dynamics for Nuclear Reactor Safety (Internet), 12 Pages, 2020/11
A bubbly flow with a single injection orifice is numerically analyzed for pool scrubbing phenomena using different computational fluid dynamics (CFD) methodologies. The calculation covers the total regime of pool scrubbing from air injection to bubble swarm through the transition region. Such two-phase flow behaviors strongly affect particle removal in a bubble. The experimental bubbles are known to be oblate spherical and exhibit secondary motion, including path instability and shape oscillations. Moreover, bubbles in a swarm are subject to coalescence and breakup. While these may well affect bubble internal heat/mass transfer and particle capture, no established way is available for considering such influences in practical calculations. Pool scrubbing code SPARC-90 uses an oblate spherical bubble model but assumes a steady, rectilinear bubble rise without secondary motion. The 3-D CFD has the potential to capture the bubble interaction in the swarm region in detail. In the present study, the experiment by Abe et al. (Nuclear Engineering and Design 337, 2018) was referred for the calculation, and their data were used to validate if the CFD simulation can predict the flow transition accurately. Two types of solvers based on the volume of fluid (VOF) method and the simple coupled volume of fluid with level set (S-CLSVOF) method are used for the interface capture. The two solvers were validated by comparing with the experimental results. As a result, the void fraction profiles along the vertical central axis were in good agreement with the experimental data, regardless of the solvers, and those along horizontal lines in a central plane slightly improved with the S-CLSVOF method by the more accurate calculation of the surface tension.
孫 昊旻; 柴本 泰照; 岡垣 百合亜; 与能本 泰介
Science and Technology of Nuclear Installations, 2019, p.1743982_1 - 1743982_15, 2019/06
被引用回数：13 パーセンタイル：82.61(Nuclear Science & Technology)Because a pool scrubbing is important for reducing radioactive aerosols to the environment for a nuclear reactor in a severe accident situation, many researches have been performed. However, decontamination factor (DF) dependence on aerosol concentration was seldom considered. DF dependence in the pool scrubbing with 2.4 m water submergence was investigated by light scattering aerosol spectrometers. It was observed that DF increased monotonically as decreasing particle number concentration in a constant thermohydraulic condition. Two validation experiments were conducted to confirm the observed DF dependence. In addition, characteristics of the DF dependence in different water submergences were investigated experimentally. It was found the DF dependence became more significant in higher water submergence.
孫 昊旻; 町田 真一*; 柴本 泰照; 岡垣 百合亜; 与能本 泰介
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07
Pool scrubbing is one of the efficient filters with a high decontamination factor (DF). Because of its high performance, many pool scrubbing research have been performed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption. In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was a real phenomenon of the pool scrubbing.
与能本 泰介; 柴本 泰照; 佐藤 聡; 岡垣 百合亜
Journal of Nuclear Science and Technology, 53(9), p.1342 - 1352, 2016/09
被引用回数：3 パーセンタイル：28.28(Nuclear Science & Technology)BWRの運転時の異常な過渡変化を超える過渡事象におけるドライアウトした燃料表面のリウェット挙動に関して、当研究グループで以前実施した研究では、リウェット直前の冷却として定義する先行冷却により、その伝播速度が強く支配されることが示された。本研究では、この先行冷却の特徴を把握するために、実験結果に対して、さらに工学解析と熱伝導解析を実施した。特徴把握のため、まず、先行冷却を熱伝達率評価値を用いて定量的に定義し、先行冷却が開始するタイミングでの被覆管温度の関数としてリウェット速度を検討した。その結果、リウェット点近傍での最大伝熱量によりリウェット速度が制限される傾向が示され、熱伝導解析の結果と整合した。
孫 昊旻; 柴本 泰照; 岡垣 百合亜; 与能本 泰介
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 8 Pages, 2016/06
Pool scrubbing is one of the effective mechanisms to filter out radioactive aerosols in a severe accident of a nuclear reactor. A lot of work has been done on the pool scrubbing models and experiments. However, large discrepancies still exist between the simulation and experimental results. To improve the pool scrubbing model, an accurate decontamination factor (DF) evaluation by an aerosol measurement is inevitable. A pool scrubbing experimental apparatus was constructed. The aerosol particle diameter distribution was measured by a light scattering aerosol spectrometer. We focused on investigating and reducing the error of DF experimentally. Several problems resulting in the error and their solutions for the error reduction were summarized in this paper. Based on the error reduction methods, the DFs of pool scrubbing were measured in two water submergences.
与能本 泰介; 柴本 泰照; 竹田 武司; 佐藤 聡; 石垣 将宏; 安部 諭; 岡垣 百合亜; 孫 昊旻; 栃尾 大輔
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5341 - 5352, 2015/08
This paper summarizes thermal-hydraulic (T/H) safety studies being conducted at JAEA based on the consideration of research issues after the Fukushima Dai-Ichi Nuclear Power Station accident. New researches have been initiated after the accident, which are related to containment thermal hydraulics and accident management (AM) measures for the prevention of core damage under severe multiple failure conditions. They are conducted in parallel with those initiated before the accident such as a research on scaling and uncertainty of the T/H phenomena which are important for the code validation. Those experimental studies are to obtain better understandings on the phenomena and establish databases for the validation of both lumped parameter (LP) and computational fluid dynamics (CFD) codes. The research project on containment thermal hydraulics is called the ROSA-SA project and investigates phenomena related to over-temperature containment damage, hydrogen risk and fission product (FP) transport. For this project, we have designed a large-scale containment vessel test facility called CIGMA (Containment InteGral Measurement Apparatus), which is characterized by the capability of conducting high-temperature experiments as well as those on hydrogen risk with CFD-grade instrumentation of high space resolution. This paper describes the plans for those researches and results obtained so far.
上地 優; 寺田 敦彦; 岡垣 百合亜; 日野 竜太郎
Journal of Nuclear Science and Technology, 51(7-8), p.964 - 967, 2014/07
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)福島第一原子力発電所の使用済セシウム吸着塔の安全な保管に向けて、水素対策は重要な課題である。吸着塔内では、残留水の放射線水分解により、水素が発生し、容器内に蓄積する可能性がある。本稿では、FLUENTコードを用いた熱流動解析結果の検証のために、模擬容器内の流動をPIVにより計測した結果を報告する。試験結果から、容器内では非常に低速な上昇流が生じるとともに、ステップ部において循環流が生じることが明らかになった。この流動様相は解析結果とも一致した。加えて、触媒による水素再結合反応についてその効果を確認した。実機環境において想定される高湿度環境においても触媒が機能することが明らかになった。
山岸 功; 永石 隆二; 寺田 敦彦; 上地 優; 加藤 千明; 森田 圭介; 西原 健司; 津幡 靖宏; Ji, W.*; 福島 久志*; et al.
IAEA-CN-211 (Internet), 5 Pages, 2013/01
福島第一原子力発電所事故では、損傷炉心の冷却のために大量の放射性汚染水が発生した。塩水を含む汚染水処理にはゼオライトによるセシウム吸着が適用されたが、使用済ゼオライトは高放射能廃棄物となるため、その安定保管が喫緊の課題である。原子力機構は水処理設備の運転で生じる二次廃棄物管理の研究開発を実施しており、使用済ゼオライトの安定保管に関しては、性状の調査及び現行保管方法の有効性評価を行っている。評価においては、崩壊熱の影響、水の放射線分解による水素発生、塩水に曝された容器の健全性を考慮する必要がある。本発表では、これまでに得られた成果について報告する。
岡垣 百合亜*; 杉山 均*; 加藤 直人*; 日野 竜太郎
自動車技術会論文集, 43(4), p.949 - 955, 2012/07
ブロック型高温ガス炉の経済性向上にかかわる燃料棒の発熱密度を高めるため、燃料棒の黒鉛シースに周期配列突起(リブ)を設けて燃料棒の乱流熱伝達を促進させることが考えられている。リブ加工時の製作性を考慮して、リブ断面は直方体形状を候補としているが、熱伝達性能に及ぼす配列のピッチやリブ高さなどの影響及び効果をあらかじめ把握しておく必要がある。そこで、今後の燃料棒設計への適用を見通して、非等方性乱流にも適用でき、しかも、解析負荷を抑制できる乱流モデルとして代数レイノルズ応力モデルに注目し、同モデルを用いて周期配列リブ付き直線管路内の乱流解析を行った。その結果、リブ付き流路に特徴的な流れの再付着点位置やリブ周辺の渦をよく再現し、同モデルはリブに起因する非等方性乱流場の予測に適用できることがわかった。
岡垣 百合亜*; 杉山 均*; 加藤 直人*; 寺田 敦彦; 日野 竜太郎
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10
高温のヘリウムガス熱交換器の伝熱促進等に有用な45度に傾斜した矩形リブを配した矩形ダクト内の乱流熱伝達特性について、新たに開発した代数応力モデルを用いて数値解析を実施した。解析結果は、PIVによる高レイノルズ数の乱流場における実験結果をよく再現しており、解析モデルの妥当性を確認できた。
岡垣 百合亜*; 杉山 均*; 日野 竜太郎; 加藤 直人*; 寺田 敦彦
日本機械学会関東支部ブロック合同講演会講演論文集, p.35 - 36, 2011/09
伝熱性能を向上させるため、壁面にリブと呼ばれる突起を周期的に配置することにより、乱れを増加させる手法が広く用いられている。今後の発展が期待される高温ガス炉(HTGR)では、水素製造プロセスなどへの効率的な核熱供給を行うため、リブによる伝熱促進を図ったプロセス熱交換器を用いる計画である。本研究は、熱交換器の設計コード開発の第一段階として、考案した代数レイノルズ応力モデルを用いて周期配列突起付き直線管路の乱流解析を行い、その妥当性を検証するため、CasarsaらのPIV計測実験結果と比較検討した。その結果、速度場,レイノルズ応力分布の実験結果との比較から、速度分布の結果は実験とほぼ一致するが、レイノルズ応力においては他の乱流モデルと同様に値や分布に違いがみられることがわかった。
寺田 敦彦; 上地 優; 日野 竜太郎; 西原 健司; 永石 隆二; 山岸 功; 岡垣 百合亜*
no journal, ,
原子力機構では、東京電力福島第一原子力発電所における高濃度汚染滞留水の浄化処理に用いられたゼオライト吸着塔(KURION)の長期保管方策の検討を進めており、その一環として吸着塔内の水素拡散解析評価を行い、ゼオライト層が乾燥状態を仮定した場合、水素濃度が爆発下限界以下であること、及び塔内温度が水素の自己着火温度以下であることを確認した。
岡垣 百合亜
no journal, ,
熱水力安全研究グループでは、原子力規制委員会による安全規制を技術的に支援することを目的に、シビアアクシデント時の格納容器内の熱流動及びソースターム移行挙動に関する実験、国産熱水力コードの整備を支援するための模擬炉心等機器での熱水力に関する実験、福島第一原子力発電所事故を踏まえた事故対策や新型安全系に関する大型非定常実験装置(LSTF)を用いた実験等を実施する。あわせて、現在、国産の最適評価(BE)コード並びに数値流体力学(CFD)コードに適用するための評価モデル等の検証・開発を進めている。今後は、実験において装置の製作・整備を進めるとともに、順次実験を開始し、解析結果を用いて妥当性評価(V&V)を行い、三次元二相流解析コードの開発・整備に反映させる。
