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Effects of nozzle orifice shape on jet breakup and splashing during liquid jet impact onto a horizontal plate

Sun, G.*; Zhan, Y.*; 大川 富雄*; 青柳 光裕; 内堀 昭寛; 岡野 靖

Experimental Thermal and Fluid Science, 151, p.111095_1 - 111095_15, 2024/02

 被引用回数:1 パーセンタイル:0.01(Thermodynamics)

Experiments were carried out on the liquid jet ejected from oval nozzles to investigate the effects of nozzle orifice shape on jet behavior. At low and high liquid flow rates, the liquid jet behaved similarly to the circular jet in our previous studies; the jet breakup lengths and the size of the droplets formed after the jet breakup were expressed by similar dimensionless correlations as those for the circular jet. At the intermediate liquid flow rate, a bamboo leaf-like structure formed on the liquid jet dominated the jet breakup. The jet breakup lengths were therefore correlated using a theory for the surface tension-induced shape oscillation of elliptical fluid. These correlations enabled to estimate the liquid jet state at any distance from the nozzle. It was also confirmed that if the state of the liquid jet at the impact point is known, the splash rate and the size of the splashed droplets can be predicted satisfactorily using the available correlations based on the experimental data for the circular jet.


Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

内堀 昭寛; 堂田 哲広; 青柳 光裕; 曽根原 正晃; 曽我部 丞司; 岡野 靖; 高田 孝*; 田中 正暁; 江沼 康弘; 若井 隆純; et al.

Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11

 被引用回数:1 パーセンタイル:72.91(Nuclear Science & Technology)



部門設立30周年記念出版Vol.3(ナトリウム冷却高速炉の開発; 社会実装に向けた熱流動・安全性研究)

田中 正暁; 内堀 昭寛; 岡野 靖; 横山 賢治; 上羽 智之; 江沼 康弘; 若井 隆純; 浅山 泰

第27回動力・エネルギー技術シンポジウム講演論文集(インターネット), 5 Pages, 2023/09

日本機械学会動力エネルギーシステム部門の30周年を記念し、「JSME Series in Thermal and Nuclear Power Generation(JSMEシリーズ 火力・原子力発電)」の第3巻として「Sodium-cooled Fast Reactor(ナトリウム冷却高速炉)」(本書)が発刊となった。本報では、本書の第5章にまとめられている、SFR開発に必要な枢要技術である熱流動及び安全性に関連するR&D成果等について概説するとともに、経験を含めた豊富な知識(ナレッジ)を活用し、最新の数値シミュレーション技術を組み合わせた革新炉の社会実装を支援する統合評価手法「ARKADIA」の開発状況について概説する。


Development of Lagrangian particle method for temperature distribution formed by sodium-water reaction in a tube bundle system

小坂 亘; 内堀 昭寛; 岡野 靖; 柳沢 秀樹*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08



The Development of a Multiphysics Coupled Solver for Studying the Effect of Dynamic Heterogeneous Configuration on Particulate Debris Bed Criticality and Cooling Characteristics

Li, C.-Y.; Wang, K.*; 内堀 昭寛; 岡野 靖; Pellegrini, M.*; Erkan, N.*; 高田 孝*; 岡本 孝司*

Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07

 被引用回数:1 パーセンタイル:0(Chemistry, Multidisciplinary)

For a sodium-cooled fast reactor, the capability for stable cooling and avoiding re-criticality on the debris bed is essential for achieving in-vessel retention when severe accidents occur. However, an unexploited uncertainty still existed regarding the compound effect of the heterogeneous configuration and dynamic particle redistribution for the debris bed's criticality and cooling safety assessment. Therefore, this research aims to develop a numerical tool for investigating the effects of the different transformations of the heterogeneous configurations on the debris bed's criticality/cooling assessment. Based on the newly proposed methodology in this research, via integrating the Discrete Element Method (DEM) with Computational Fluid Dynamics (CFD) and Monte-Carlo-based Neutronics (MCN), the coupled CFD-DEM-MCN solver was constructed with the originally created interface to integrate two existing codes. The effects of the different bed configurations' transformations on the bed safety assessments were also quantitively confirmed, indicating that the effect of the particle-centralized fissile material had the dominant negative effect on the safety margin of avoiding re-criticality and particle re-melting accidents and had a more evident impact than the net bed-centralized effect. This coupled solver can serve to further assess the debris bed's safety via a multi-physics simulation approach, leading to safer SFR design concepts.


Development of the ex-vessel modules for the integrated SFR safety analysis code SPECTRA

青柳 光裕; 牧野 徹*; 大木 裕*; 内堀 昭寛; 岡野 靖

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05

The SPECTRA code has been developed as an integrated safety analysis tool for sodium-cooled fast reactors (SFRs). In this study, the capability of SPECTRA is enhanced by integrating the analyses of sodium pool fire and concrete ablation for overlapped events of the ex-vessel phenomena. The sodium pool fire module is connected to the shared module for the sodium pool and the floor concrete developed in our previous study. The developed model is validated through the benchmark analysis of the F7-1 pool fire experiment. The calculation result of the pool and catch pan temperature shows good agreement with the experimental data. A demonstration analysis is also conducted for an overlapped event of the ex-vessel phenomena.


Development of dynamic PRA methodology for external hazards in sodium-cooled fast reactor via applying Markov chain Monte Carlo method to severe accident analysis code; Assessment of accident management of assigning independent emergency diesel generators to each air cooler

Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

Quantitative assessment of the effect of accident management on the various external hazards is essential in the nuclear safety analysis. This study aims to establish the dynamic probabilistic risk assessment methodology for sodium-cooled fast reactors that can consider the transient plant status under continuous external hazards with corresponding countermeasures operating stochastically. Specifically, the Continuous Markov chain Monte Carlo (CMMC) and Deterministic and Stochastic Petri Nets (DSPN) methods are newly applied to the severe accident analysis code, SPECTRA, which can conduct dynamic plant evaluation in the different severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. In the DSPN-CMMC-SPECTRA coupled frame, the latest safety functions of the plant components/systems can be stochastically determined by the DSPN-CMMC grounded on the current plant states under continuous hazard and the interaction between the multi-state components/systems; then, SPECTRA can evaluate the following plant state determined by the latest safety function of the components/systems. Therefore, the advantage of this newly developed DSPN-CMMC-SPECTRA frame is having the capability to quantitatively and stochastically evaluate the transient accident progressions that potentially lead to the core damage under the continuous external hazard scenario. As for the preliminary exam on the DSPN-CMMC-SPECTRA frame, one of the typical external hazards of continuous volcanic ashfall is selected in this research. In addition, the numerical investigation of alternative accident management' effects has also been carried out and quantitatively confirmed in this research.


Experimental study on the breakup of liquid jet discharged from a nozzle with sudden contraction

Sun, G.*; 大川 富雄*; 青柳 光裕; 内堀 昭寛; 岡野 靖

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05



ARKADIA; For the innovation of advanced nuclear reactor design

大島 宏之; 浅山 泰; 古川 智弘; 田中 正暁; 内堀 昭寛; 高田 孝; 関 暁之; 江沼 康弘

Journal of Nuclear Engineering and Radiation Science, 9(2), p.025001_1 - 025001_12, 2023/04



炉内ソースターム解析コードTRACER Version 2.4.1(マニュアル)

大野 雅広*; 内堀 昭寛; 岡野 靖; 高田 孝*

JAEA-Testing 2022-004, 193 Pages, 2023/03


高速炉の燃料破損時にナトリウム中に放出される放射性物質の挙動は、燃料破損の速やかな検出によるプラント異常事象の拡大防止、保守時の被曝線量の低減、及び事故時に放出される放射性物質量評価等に関して重要である。このため、燃料破損時に冷却材中に放出され、一次冷却材を経由してカバーガス空間へ至る核分裂生成物(以下、FPと略す)の種類とその量(炉内ソースターム)をより現実的に評価することを目的として、これらの FP 移行過程で起こる物理的・化学的挙動を機構論的に取り扱う解析コードTRACER (Transport phenomena of Radionuclides for Accident Consequence Evaluation of Reactor)が開発されている。TRACERコードは、燃料ピンの破損に伴う冷却材へのFP放出から始まる、一連のFP移行挙動を解析する。解析は燃料ピン、一次冷却材及びカバーガスと炉内の範囲でのFP挙動を対象としている。具体的には、燃料ピンから放出されるFP、1次系冷却材中を移行するFP、冷却材中を輸送されるFPを含む希ガス気泡、カバーガスへ放出されるFP、カバーガスから炉外へ漏洩するFPといった一連の挙動である。本マニュアルはTRACER Version 2.3のマニュアルに対し、数式等の参考文献の追加、インプットファイル作成方法の解説の改善、TRACERコードへ加えたNUREG-0772モデルの改良に関して追記、Appendixで行ったサンプル解析の図の修正、サンプル解析の追加といった変更を加えたものである。


The Development of Petri Net-based continuous Markov Chain Monte Carlo methodology applying to dynamic probability risk assessment for multi-state resilience systems with repairable multi-component interdependency under longtermly thereat

Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖

Journal of Nuclear Science and Technology, 23 Pages, 2023/00

 被引用回数:1 パーセンタイル:72.91(Nuclear Science & Technology)

For all the nuclear reactor systems, quantitative assessment of the accident management (AM) effects against long-term external hazards became one of the essential issues after the lesson learned from the Fukushima Daiichi Nuclear Power Plant accident. However, the influence from the safety systems' stochastic and dynamic shifting between multiple working states, which is related to the interaction with the adjacent components/systems in general, has not been accounted for yet. Therefore, this research aims to develop a dynamic probability risk assessment tool considering repairable multi-component interdependency for investigating the AM influences on the multi-state safety systems under long-term external hazards. Based on the newly proposed methodology in this research via integrating the Petri Net (PN) model with the continuous Markov chain Monte Carlo (CMMC) method, a framework applying PN-CMMC methodology to a severe accident analysis code, SPECTRA, had been originally constructed. Different AM influences on the multi-state decay heat removal systems against long-term volcanic ashfall were also quantitatively confirmed, indicating that halving the repairing time is more influential in suppressing the core damage frequency than doubling the number of adjacent electricity support systems. Therefore, the PN-CMMC-SPECTRA framework can further assess the uncharted dynamic multi-state concerns, leading to a safer AM strategy.


Development of ARKADIA-Safety for severe accident evaluation of sodium-cooled fast reactors

青柳 光裕; 曽根原 正晃; 石田 真也; 内堀 昭寛; 川田 賢一; 岡野 靖; 高田 孝

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09

Development of Advanced Reactor Knowledge- and Artificial Intelligence (AI)-aided Design Integration Approach through the whole plant lifecycle (ARKADIA) has been started in Japan Atomic Energy Agency. ARKADIA can automatically provide possible solutions of design, safety measures, and a maintenance program to optimize the lifecycle performance of advanced reactors by using the state-of-the-art numerical simulation technologies. In the first phase of this project, ARKADIA-Safety is developed for the purpose of automatic optimization of the severe accident (SA) management and its feedback to the plant design of sodium-cooled fast reactors (SFRs). This paper describes the overview of ARKADIA-Safety and its application for SA evaluation.


Validation study of sodium pool fire modeling efforts in MELCOR and SPHINCS codes

Louie, D. L. Y.*; 青柳 光裕; 内堀 昭寛; 高田 孝; Luxat, D. L.*

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 6 Pages, 2022/09

The paper describes progress of an international collaborative research in the area of SFR sodium fire modeling between the United States and Japan under the framework of the Civil Nuclear Energy Research and Development Working Group (CNWG). In this collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA), the validation basis for and modeling capabilities of sodium spray and pool fires in MELCOR of SNL and SPHINCS of JAEA are being enhanced. This study documents MELCOR and SPHINCS sodium pool fire model validation exercises against the JAEA's sodium pool fire experiments, F7-1 and F7-2. The proposed enhancement of the sodium pool fire models in MELCOR through addition of thermal hydraulic and sodium spreading models that enable a better representation of experimental results is also described.


Development of plant lifecycle optimization method, ARKADIA for advanced reactors

内堀 昭寛; 曽我部 丞司; 岡野 靖; 高田 孝*; 堂田 哲広; 田中 正暁; 江沼 康弘; 若井 隆純; 浅山 泰; 大島 宏之

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 10 Pages, 2022/09



Experiment study on the effect of nozzle shape on liquid jet breakup

Sun, G.*; Zhan, Y.*; 大川 富雄*; 青柳 光裕; 内堀 昭寛; 岡野 靖

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08



Development of dynamic PRA methodology for external hazards (Application of CMMC method to severe accident analysis code)

Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖

第26回動力・エネルギー技術シンポジウム講演論文集(インターネット), 4 Pages, 2022/07



Sodium-cooled Fast Reactors

大島 宏之; 森下 正樹*; 相澤 康介; 安藤 勝訓; 芦田 貴志; 近澤 佳隆; 堂田 哲広; 江沼 康弘; 江連 俊樹; 深野 義隆; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

ナトリウム冷却高速炉(SFR: Sodium-cooled Fast Reactor)の歴史や、利点、課題を踏まえた安全性、設計、運用、メンテナンスなどについて解説する。AIを利用した設計手法など、SFRの実用化に向けた設計や研究開発についても述べる。


A Preliminary validation study for removal performance of iodine gas in sodium pool with a simplified approach

Kam, D. H.*; Grabaskas, D.*; Starkus, T.*; Bucknor, M.*; 内堀 昭寛

Transactions of the American Nuclear Society, 126(1), p.536 - 539, 2022/06

ナトリウム冷却高速炉の燃料ピン破損事故を評価する上で、破損燃料から放出された気泡に含まれる放射性物質の周囲ナトリウムへの移行挙動が重要となる。アルゴンヌ国立研究所におけるソースターム挙動解析コードSRT(Simplified Radionuclide Transport)の整備の一環として、気泡中放射性物質の移行挙動のモデル化に資するため、ナトリウム中ヨウ素ガス移行試験の数値解析を実施した。これにより、提案する評価手法が計測された除染係数を概ね再現できることを確認した。


Development of integrated severe accident analysis code, SPECTRA for sodium-cooled fast reactor

内堀 昭寛; 曽根原 正晃; 青柳 光裕; 高田 孝*; 大島 宏之

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04



Development of the sodium pool and floor concrete module for the integrated SFR safety analysis code, SPECTRA

青柳 光裕; 内堀 昭寛; 高田 孝

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03

The SPECTRA code has been developed as an integrated safety analysis tool for sodium-cooled fast reactors (SFRs). In this study, the capability of SPECTRA is enhanced by establishing a sodium pool and floor concrete module commonly used in individual physical modules. This paper describes the framework of modelling for the mass and heat transfer in the sodium pool and the floor concrete. Considering concrete ablation due to sodium and debris, bottom of the sodium pool changes during event progression. This change in material composition in certain position is modeled by volume and mass fraction of each component. A simple convection model for the pool is implemented to ensure the conservation of heat and mass. This model is tested through the verification analysis in comparison with the existing model. As the result, it is confirmed the behavior of pool spreading and concrete ablation can be simulated by this module correctly.

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