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田中 正暁; 浜瀬 枝里菜; Li, S.*; 遠藤 淳二*; Gerschenfeld, A.*; Gorsse, Y.*; 今村 亮*; 今村 友*; 吉村 一夫; 江連 俊樹
Proceedings of 10th Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-10) (Internet), 16 Pages, 2025/00
ナトリウム冷却高速炉(SFR)の安全性向上のため、多様な崩壊熱除去システム(DHRS)が必要である。原子炉容器内のホットプレナムに浸漬型直接熱交換器(D-DHX)を備えた直接原子炉補助冷却システムは、通常運転条件だけでなく、予想される運転事象および事故条件においても最も有望なシステムの一つであると認識されている。本論文では、D-DHX運転中に発生する複雑な熱流動現象下での炉心冷却性を評価するための数値流体力学(CFD)モデリングおよびシミュレーション技術の統合を試みた。炉心の燃料集合体およびラッパー間ギャップを含む原子炉容器内の適切なCFD解析モデルに関するコンセンサスを得ることを目的に、原子力機構のナトリウム試験施設(PLANDTL-1)、CEAの革新的SFRであるASTRID、および日本のプール型実証SFRに関するMFBRのCFD解析に関するそれぞれの経験のレビューとギャップ分析を行った。
岡垣 百合亜; 竹田 武司; 和田 裕貴; 安部 諭; 市原 京子*; 塩谷 仁*
Proceedings of 10th Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-10) (Internet), 12 Pages, 2025/00
Core Exit Temperature (CET), measured by core exit thermocouples (T/Cs), is utilized worldwide as a crucial parameter to start Accident Management (AM) operator action by detecting core temperature excursion during accidents in Pressurized Water Reactors (PWRs). The CET is used to switch accident response procedures from preventing core damage to preventing containment failure. Various thermal-hydraulic phenomena in the core influence the behavior of CET during an accident. Previous studies have indicated that CET may rise more slowly than the fuel cladding temperature. This study aimed to deepen the understanding of the relationship between CET and fuel cladding temperature during representative accident progressions by employing Computational Fluid Dynamics (CFD) simulations. It sought to complement experimental findings by evaluating the effects of three-dimensional core thermo-hydraulic behaviors, such as secondary flows in the upper core region. The CFD simulation was performed for a 1% vessel upper head small-break Loss-Of-Coolant Accident (LOCA) experiment, which was conducted at the Large-Scale Test Facility (LSTF) at Japan Atomic Energy Agency (JAEA) in 2023. The LSTF experiment assumed that the high-pressure injection system of the emergency core cooling system had totally failed. This study focused on the representative period when a significant rise in core temperature appeared during core uncovering. The transient solver for compressible fluids in OpenFOAM was employed for the CFD simulation. Boundary conditions, such as mass flow rate, temperature, and pressure at the core's top position, were applied. The turbulence model used was the Shear Stress Transport (SST) 
model. The CET distributions were compared with the experimental data, which had a total of 20 points. The velocity and temperature distributions in the mainstream and cross-sectional directions were visualized to elucidate thermal-hydraulic phenomena. This study provided valuable insights into CET behavior and related thermo-fluid dynamics.
浜瀬 枝里菜; 藤崎 竜也*; 河村 拓己*; 三宅 康洋*; 堂田 哲広; 小野 綾子; 田中 正暁
Proceedings of 10th Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-10) (Internet), 12 Pages, 2025/00
浸漬型直接炉内冷却器を用いた自然循環崩壊熱除去時のプラント全体挙動を精度よく評価するため、プラント動特性解析コードとCFDコードを用いた連成解析手法を整備している。PLANDTL-1のPRACSを用いたスクラム過渡模擬試験を対象に、連成解析を実施した結果、原子炉容器(RV)内の局所的な現象を考慮して、系統内熱流動挙動を評価するとともに、本挙動をフィードバックしながら、RV内熱流動現象を模擬することができ、本手法の適用性を確認できた。
小坂 亘; 内堀 昭寛; 渡部 晃*; 椎名 祥己*
Proceedings of 10th Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-10) (Internet), 12 Pages, 2025/00
In a steam generator (SG) of a sodium-cooled fast reactor, water leakage caused by heat transfer tube failure in an SG makes a high-temperature, high-velocity, and corrosive jet with sodium-water reaction (SWR). The reacting jet may cause further tube failure and the expansion of the affected area in an SG. A mechanistic analysis code SERAPHIM has capability to evaluate the reacting jet considering multi-component, multiphase and compressive flow with SWR. By comparing analysis and experimental results, SERAPHIM code has been validated from the case of a free jet to the reacting jet in a tube bundle system with an average water leak rate of 0.15 kg/s. In this paper, as one of the validation analysis series of SERAPHIM code, we performed numerical analysis for the reacting jet in a tube bundle system with larger average water leak rate, i.e. 1.85 kg/s. Similar to previously reported results, the gas phase expanded in the system immediately after the discharging and then flowed in the direction of buoyancy force. It was confirmed the occurrence of an underexpanded jet and the limitation of the maximum temperature, which are consistent with the knowledge of the SWR phenomena in an SG. The low and high-temperature regions were wider than in previous reports due to the larger average water leak rate. The high-temperature region obtained by the analysis agreed with the representative characteristics appeared in the experimental result such as location and shape of the temperature distribution. The analysis results were consistent with previous knowledge. The representative characteristics that appeared in the experimental result were also reproduced. SERAPHIM code was applicable to the selected average water leak rate in the tube bundle system.