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Development of a formulation to predict molten core spreading in an LWR severe accident

Sahboun, N. F.; 松本 俊慶; 岩澤 譲; Wang, Z.; 杉山 智之

Annals of Nuclear Energy, 195, p.110145_1 - 110145_12, 2024/01

Relocated corium into the Primary Containment Vessel needs to be properly cooled to avoid or mitigate molten core concrete interactions in the PCV in order to maintain its supporting capability for the reactor pressure vessel and to suppress combustible or non-condensable gas releases. To know how effective the cooling is, it became important to know the geometry of the relocated corium. The present study chooses to focus on the "Wet Cavity" strategy and to build a reliable tool to evaluate the corium coolability in such a case. To achieve this goal, a previously developed formulation built to predict the corium geometry under the "Dry Cavity" strategy was extended to the conditions used in the "Wet Cavity" strategy. This extension includes the effect of solidification and cooling from the water by using a newly developed expression for the dimensionless thickness s, the water subcooling, and the melts super heat. After the validation of the extended formulation was confirmed, potential restrictions and limitations were investigated.


High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 三原 武; 垣内 一雄; 宇田川 豊

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5$$^{TM}$$ cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.


Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

浜瀬 枝里菜; 大釜 和也; 河村 拓己*; 堂田 哲広; 田中 正暁; 山野 秀将

Annals of Nuclear Energy, 195, p.110157_1 - 110157_14, 2024/01

高速炉プラント動特性解析コードSuper-COPDのスクラム不作動流量喪失事象に対する妥当性確認のため、FFTFの受動的安全性試験LOFWOS No.13試験を対象としたIAEAベンチマークに参加した。ブラインドフェーズで課題として抽出された燃料集合体出口温度及び全反応度の評価精度向上のため、集合体間熱移行及び集合体間ギャップ部流れを考慮した全炉心モデル及び炉心湾曲反応度簡易評価モデルを導入した。最終フェーズ解析の結果、2次ピーク時の集合体出口温度を良好に再現するとともに、全反応度の実測値の挙動を概ね評価できたことから、LOFWOSに対するSuper-COPDの妥当性を確認した。


Experiment and new analysis model simulating in-place cooling of a degraded core in severe accidents of sodium-cooled fast reactors

今泉 悠也; 青柳 光裕; 神山 健司; 松場 賢一; Akaev, A.*; Mikisha, A.*; Baklanov, V.*; Vurim, A.*

Annals of Nuclear Energy, 194, p.110107_1 - 110107_11, 2023/12

In severe accidents of SFRs, the cooling of the residual core materials, which is called "in-place cooling", is one of the important factors for In-Vessel Retention (IVR). For the evaluation, behavior of the in-place cooling was analyzed by the SIMMER-III code. In order to understand the in-place cooling, method of Phenomena Identification and Ranking Table (PIRT) was applied. Based on the result, an out-of-pile experiment which focused on the extracted factors was conducted. In the experiment, continuous oscillation of sodium level was observed by sodium vaporization and condensation. Analysis for the out-of-pile experiment was conducted by SIMMER-III, but the results were different between the experiment and the analysis. By investigation of the analysis result, it was revealed that the difference was due to occupation of non-condensable gas. Therefore, an analysis model of inter-cell gas mixing was newly developed, and the agreement was significantly improved by the new model.


Scalability of inertial particle deposition in bubbles with internal circulation

茂木 孝介; 柴本 泰照; 久木田 豊

Annals of Nuclear Energy, 184, p.109679_1 - 109679_10, 2023/05

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)

Inertial deposition of small (less than a few $$mu$$ m in diameter) aerosol particles in mm-scale bubbles is an old but unsettled issue in modeling of pool scrubbing phenomenon. Whereas existing practical models give no specific consideration to the bubble-internal transport, some studies have shown that inertial transport affects considerably the particle deposition rate. We show, on the basis of Lagrangian simulations of particles advected by steady internal circulation in a spherical bubble, that particle centrifugal velocity becomes scale invariant for low- Stokes numbers (St $$le$$ $$10^{-2}$$) when the characteristic timescale is chosen to be that for transversal particle motion at the Stokes terminal velocity corresponding to the local fluid acceleration. Because a scaling law can be derived by running simulations with a small number of particles, it can provide a practical tool for considering the influence of inertial particle transport within the bubble on the decontamination factor.


LASSO reconstruction scheme for radioactive source distributions inside reactor building rooms with spectral information and multi-radionuclide contaminated situations

Shi, W.*; 町田 昌彦; 山田 進; 吉田 亨*; 長谷川 幸弘*; 岡本 孝司*

Annals of Nuclear Energy, 184, p.109686_1 - 109686_12, 2023/05

 被引用回数:1 パーセンタイル:79.09(Nuclear Science & Technology)

Clarification of radioactive source distributions is one of the most important steps in initial decommissioning of not only normally shutdown reactors but also damaged ones by accidents like Fukushima Daiichi Nuclear Power Plants (FDNPP). Generally, since radioactive hot spots are restricted into specific areas in normal operating conditions, the clarification scheme can be mapped onto the inverse estimation in sparse source distributions. On the other hand, the fact that radioactive hot spots are largely spread in unknown manner as seen in FDNPP motivates to construct an inversion scheme in non-sparse source conditions. Thus, a reconstruction scheme applicable to both sparse and non-sparse radioactive distributions is highly in demand. In addition, a variety of radionuclides is produced in reactors. Thus, we also need a scheme to distinguish each source distribution in mixed multi radionuclides. In this paper, we confirm that the inverse estimation scheme using Least Absolute Shrinkage and Selection Operator (LASSO) method with spectral information commonly shows excellent performance in the above all situations. The proposed LASSO scheme with the spectral information enables to reduce the number of measurement points in sparse conditions, while information proliferation by sensing the spectrum makes it possible to directly reconstruct source distribution as almost solvable problems in non-sparse ones. Moreover, the LASSO scheme allows to reconstruct the source distribution of each potential radionuclide in multi-radionuclide coexisting situations. Consequently, we confirm that the LASSO scheme to reconstruct radioactive sources is promising for the future nuclear decommissioning projects widely from normally shutdown reactors to damaged ones like FDNPP.


Feasibility study on reprocessing of HTGR spent fuel by existing PUREX plant and technology

深谷 裕司; 後藤 実; 大橋 弘史

Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



A Quantitative method of eutectic reaction study between boron carbide and stainless steel

Hong, Z.*; Pellegrini, M.*; Erkan, N.*; Liao, H.*; Yang, H.*; 山野 秀将; 岡本 孝司*

Annals of Nuclear Energy, 180, p.109462_1 - 109462_9, 2023/01

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



Improvement of JASMINE code for ex-vessel molten core coolability in BWR

松本 俊慶; 川部 隆平*; 岩澤 譲; 杉山 智之; 丸山 結

Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12



A 3D particle-based simulation of heat and mass transfer behavior in the EAGLE ID1 in-pile test

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:35.78(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.


Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code

Simanullang, I. L.*; 中川 直樹*; Ho, H. Q.; 長住 達; 石塚 悦男; 飯垣 和彦; 藤本 望*

Annals of Nuclear Energy, 177, p.109314_1 - 109314_8, 2022/11

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)

Power distribution plays a significant role in preventing the fuel temperature exceeds the safety limit of 1600$$^{circ}$$C in high-temperature gas-cooled reactors. The experiment to measure the power distribution in the graphite-moderated system was carried out with the Very High Temperature Reactor Critical Assembly facility. In the previous study, the power distribution in the VHTRC was calculated using a nuclear design code system based on diffusion calculation. The results showed a maximum discrepancy of up to 20 between the experiment and calculated values in the axial direction. The large discrepancy occurred near the boundary of fuel and reflector regions. This study describes the evaluation results of pin-wise power distribution of the VHTRC with the Monte Carlo MVP3 code. The calculation results were in good agreement with the measured results. In the axial direction, the discrepancy was less than 1 around the boundary of fuel and reflector regions.


Clearance measurement for concrete waste generated by the decommissioning of uranium processing facilities

横山 薫; 大橋 裕介

Annals of Nuclear Energy, 175, p.109240_1 - 109240_7, 2022/09

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



BWR lower head penetration failure test focusing on eutectic melting

山下 拓哉; 佐藤 拓未; 間所 寛; 永江 勇二

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)

Decommissioning work occasioned by the Fukushima Daiichi Nuclear Power Station (1F) accident of March 2011 is in progress. Severe accident (SA) analysis, testing, and internal investigation are being used to grasp the 1F internal state. A PWR system that refers to the TMI-2 accident is typical for SA codes and testing, on the other hand, a BWR system like 1F is uncommon, understanding the 1F internal state is challenging. The present study conducted the ELSA-1 test, a test that focused on damage from eutectic melting of the liquid metal pool and control rod drive (CRD), to elucidate the lower head (LH) failure mechanism in the 1F accident. The results demonstrated that depending on the condition of the melt pool formed in the lower plenum, a factor of LH boundary failure was due to eutectic melting. In addition, the state related to the CRD structure of 1F unit 2 were estimated.


Numerical simulation of sodium mist behavior in turbulent Rayleigh-B$'e$nard convection using new developed mist models

大平 博昭*; 田中 正暁; 吉川 龍志; 江連 俊樹

Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07

 被引用回数:1 パーセンタイル:35.78(Nuclear Science & Technology)

ナトリウム冷却高速炉(SFR)のカバーガス領域におけるミスト挙動を高精度で評価するため、混合気体のレイリー・ベナール対流(RBC)に対する乱流モデルを選定するとともに、ミストに対するレイノルズ平均数密度とミストの運動量方程式を開発し、OpenFOAMコードに組み込んだ。最初に、単純な並列チャネルのRBCを、Favre平均k-$$omega$$SSTモデルを使用して計算した。その結果、平均温度と流量特性はDNS, LES、および実験の結果とよく一致した。次に、本乱流モデルと新しく開発したミストモデルを用いて、SFRのカバーガス領域を模擬した熱伝達試験装置を計算した。その結果、計算された高さ方向の平均温度分布とミスト質量濃度が試験結果とよく一致した。本研究により、SFRのカバーガス領域において乱流RBC環境でのミスト挙動を高精度にシミュレートできる手法を開発した。


An Experimental study related to axial constraint of fuel rod under LOCA conditions

永瀬 文久

Annals of Nuclear Energy, 171, p.109052_1 - 109052_8, 2022/06

 被引用回数:1 パーセンタイル:59.27(Nuclear Science & Technology)



Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

垣内 一雄; 天谷 政樹; 宇田川 豊

Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06

 被引用回数:3 パーセンタイル:82.38(Nuclear Science & Technology)

In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, an irradiation growth testing in the Halden reactor of Norway was conducted on various fuel cladding materials including the improved Zr alloy. In this paper, the effect of hydrogen, which was absorbed in the cladding tube due to corrosion, on the irradiation growth behavior was evaluated. Comparison between the specimens with or without pre-charged hydrogen revealed that the effect of hydrogen absorption, accelerating irradiation growth, became significant when the hydrogen content exceeded the hydrogen solubility limit in the corresponding irradiation temperature. Analysis based on this understanding derived growth acceleration effect (0.06$$pm$$0.01)%/100 ppm, whose denominator is defined as the amount of absorbed hydrogen involved in hydride precipitation under irradiation as a relevant parameter.


Material attractiveness evaluation of fuel assembly of accelerator-driven system for nuclear security and non-proliferation

大泉 昭人; 菅原 隆徳; 相楽 洋*

Annals of Nuclear Energy, 169, p.108951_1 - 108951_9, 2022/05

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



Adjoint-weighted correlated sampling for $$k$$-eigenvalue perturbation in Monte Carlo calculation

Tuya, D.; 長家 康展

Annals of Nuclear Energy, 169, p.108919_1 - 108919_9, 2022/05

 被引用回数:1 パーセンタイル:35.78(Nuclear Science & Technology)



Problems on neutron production data of Be-9 in TENDL-2017 and -2019 deuteron sub-libraries

権 セロム*; 今野 力; 太田 雅之*; 佐藤 聡*

Annals of Nuclear Energy, 169, p.108932_1 - 108932_7, 2022/05

 被引用回数:2 パーセンタイル:59.27(Nuclear Science & Technology)



Numerical analysis of natural convection behavior in density stratification induced by external cooling of a containment vessel

石垣 将宏*; 安部 諭; Hamdani, A.; 廣瀬 意育

Annals of Nuclear Energy, 168, p.108867_1 - 108867_20, 2022/04

 被引用回数:3 パーセンタイル:73.56(Nuclear Science & Technology)

It is essential to improve computational fluid dynamics (CFD) analysis accuracy to estimate thermal flow in a containment vessel during a severe accident. Previous studies pointed out the importance of the influence of initial and boundary conditions on the CFD analysis. The purpose of this study is to evaluate the influence of initial and boundary conditions by numerical analysis of natural convection experiments in a large containment vessel test facility CIGMA(Containment InteGral effects Measurement Apparatus). A density stratification layer was initially formed in the vessel using helium and air, and external cooling of the vessel surface-induced natural convection. In this study, we carried out numerical simulations of the density stratification erosion driven by the natural convection using the RANS (Reynolds averaged Navier-Stokes) model. As a result, the temperature boundary condition of the small internal structure in the vessel had a significant influence on the fluid temperature distribution in the vessel. The erosion velocity of the density stratification layer changed depending on the initial gas concentration distribution. Then, appropriate settings of the temperature and gas concentration conditions are necessary for accurate analysis.

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