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論文

原子炉建屋の等価線形三次元FEM解析による地盤-建物連成系地震応答に関する基礎的検討

市原 義孝*; 中村 尚弘*; 飯島 国彦*; 崔 炳賢; 西田 明美

構造工学論文集,B, 68B, p.271 - 283, 2022/04

本論文は、振動数に依存しない複素減衰を用いた鉄筋コンクリートの等価線形解析法の原子炉建屋の耐震設計への適用性を評価することを目的とする。そのため、理想的な地盤条件での原子炉建屋の非線形及び等価線形地震応答に着目して、地盤-建屋連成系の三次元FEM解析を実施した。その結果、等価線形解析結果は非線形解析結果と概ねよく整合し、その有効性を明らかにした。さらに、今回の等価線形解析法は、非線形解析モデルと比較して、構造の剛性を低めに評価する傾向があった。このため、最大せん断ひずみの評価では、非線形解析の結果よりもひずみの値が大きくなる可能性が高いことに留意する必要がある。

論文

French-Japanese experimental collaboration on fuel-coolant interactions in sodium-cooled fast reactors

Johnson, M.*; Delacroix, J.*; Journeau, C.*; Brayer, C.*; Clavier, R.*; Montazel, A.*; Pluyette, E.*; 松場 賢一; 江村 優軌; 神山 健司

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

シビアアクシデントに関する日仏共同実験の一環として、ナトリウム冷却高速炉の原子炉容器内下部プレナムへ溶融燃料が流出した時の燃料-冷却材相互作用について、その解明に向けた研究を実施している。MELT施設では、ナトリウム中へ流出したキログラム単位の模擬溶融炉心物質が急冷される様子をX線で可視化することができる。現在準備中のSERUA施設では、融体と冷却材の接触境界面温度が上昇した場合の沸騰熱伝達を評価するためのデータ取得を予定している。この論文では、これらの施設を活用した実験協力の現状について紹介する。

論文

Developing accelerator mass spectrometry capabilities for anthropogenic radionuclide analysis to extend the set of oceanographic tracers

Hain, K.*; Martschini, M.*; G$"u$lce, F.*; 本多 真紀; Lachner, J.*; Kern, M.*; Pitters, J.*; Quinto, F.*; 坂口 綾*; Steier, P.*; et al.

Frontiers in Marine Science (Internet), 9, p.837515_1 - 837515_17, 2022/03

 被引用回数:0 パーセンタイル:0.02(Environmental Sciences)

Vienna Environmental Research Accelerator (VERA)における加速器質量分析(AMS)の最近の大きな進歩は、検出効率向上とアイソバー抑制向上であり、環境中の極微量の長寿命放射性核種を分析する可能性を開くものである。これらの核種は$$^{233}$$U, $$^{135}$$Cs, $$^{99}$$Tc及び$$^{90}$$Srであり、通常は安定して海水中に溶存していることから、海洋混合・循環や放射性物質の広がりを研究する海洋トレーサーへの適応が重要になる。特に、同位体比$$^{233}$$U/$$^{236}$$Uと$$^{137}$$Cs/$$^{135}$$Csは元素分別の影響を受けないため、放出源の特定に有力なフィンガープリントであることが我々の研究によって実証されている。検出効率の向上により、10Lの海水試料で主要長寿命アクチニド$$^{236}$$U, $$^{237}$$Np, $$^{239,240}$$Pu, $$^{241}$$Amに加え、非常に稀な$$^{233}$$Uを分析することが可能となり、北西太平洋におけるアクチノイドの典型的な深度プロファイルを得ることに成功した。特に$$^{90}$$Sr分析に関しては、IAEAの標準物質(例えばIAEA-A-12)を用いて我々の新しいアプローチが海洋学的研究へ応用可能であることを示した。我々の推定では、$$^{90}$$Srと$$^{135}$$Csそれぞれの分析に必要な海水はわずか(1-3)Lである。

論文

Development of dispersed phase tracking method for time-series 3-dimensional interface shape data

堀口 直樹; 吉田 啓之; 山村 聡太*; 藤原 広太*; 金子 暁子*; 阿部 豊*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 14 Pages, 2022/03

In severe accidents of nuclear reactors, molten fuel and structural materials leak out of the pressure vessel into the water pool on the pedestal floor. If the water pool is shallow, the molten material enters the shallow pool as a liquid jet, disperses as debris, spreads over the floor, and it cooled by fuel-coolant interaction (FCI). Numerical simulations and experiments with state-of-the-art visualization techniques are developed and used to consider the thermal-hydraulic behavior of the liquid jet as a debris jet. By performing these simulations and experiments, we obtain detailed 3-dimensional shapes of the liquid jet interfaces. However, to evaluate the thermal-hydraulic behavior of the liquid jet, we require not only 3-dimensional shapes but also the velocity and size of dispersed liquid. We have developed a dispersed phase tracking method by using time-series data of 3-dimensional shapes of the melt interface obtained by numerical simulations or experiments to obtain these data. Firstly, we verified the applicability of the developed method by applying a simple system. Next, we applied the method to the numerical results of a liquid jet entering a shallow pool by TPFIT. The results show that the liquid jet entering the shallow pool reproduces the dispersion behavior of the fragments. The generated fragments were quantitively confirmed to have curved and rotational trajectories with complex nonlinear motions. In the relationship between the volume equivalent diameter of the fragments and the magnitude of velocity, it was confirmed that the larger the equivalent diameter, the smaller the velocity fluctuation.

論文

3D FEM soil-structure interaction analysis for Kashiwazaki-Kariwa Nuclear Power Plant considering soil separation and sliding

市原 義孝*; 中村 尚弘*; 森谷 寛*; 崔 炳賢; 西田 明美

Frontiers in Built Environment (Internet), 7, p.676408_1 - 676408_14, 2021/06

本論文は、原子炉建屋/機器・設備の現実的応答評価の精度向上を目的に、建屋-地盤境界部の剥離・滑りを考慮した3次元FEMモデルにより2007年新潟県中越沖地震時の柏崎刈羽原子力発電所7号機原子炉建屋のシミュレーション解析を実施し、得られた知見をまとめたものである。3次元FEMモデルによる建屋-地盤連成系のシミュレーション解析から、基礎版端部で基礎浮き上がりが生じ、その影響が建屋側面及び底面の土圧性状、埋め込み部表層の最大応答加速度に局部的な応答の差異となって現れることを明らかにした。今回の検討においては、基礎浮き上がりと剥離・滑りが埋め込み部表層の最大応答加速度、建屋側面及び底面の土圧性状に与える影響は比較的小さかったものの、今後、さらに大きな地震動を想定する場合には、これらの影響が増大する可能性が考えられるため、地震応答解析においてはこれら影響の適切な評価が必要になると考えられる。

論文

Spin-orbit-induced Ising ferromagnetism at a van der Waals interface

松岡 秀樹*; Barnes, S. E.*; 家田 淳一; 前川 禎通; Bahramy, M. S.*; Saika, B. K.*; 竹田 幸治; 和達 大樹*; Wang, Y.*; 吉田 訓*; et al.

Nano Letters, 21(4), p.1807 - 1814, 2021/02

 被引用回数:6 パーセンタイル:87.05(Chemistry, Multidisciplinary)

Magnetocrystalline anisotropy, a key ingredient for establishing long-range order in a magnetic material down to the two-dimensional (2D) limit, is generally associated with spin-orbit interaction (SOI) involving a finite orbital angular momentum. Here we report strong out-of-plane magnetic anisotropy without orbital angular momentum, emerging at the interface between two different van der Waals (vdW) materials, an archetypal metallic vdW material NbSe$$_{2}$$ possessing Zeeman-type SOI and an isotropic vdW ferromagnet V$${}_5$$Se$${}_8$$. We found that the Zeeman SOI in NbSe$$_{2}$$ induces robust out-of-plane magnetic anisotropy in V$$_{5}$$Se$$_{8}$$ down to the 2D limit with a more than 2-fold enhancement of the transition temperature. We propose a simple model that takes into account the energy gain in NbSe$$_{2}$$ in contact with a ferromagnet, which naturally explains our observations. Our results demonstrate a conceptually new magnetic proximity effect at the vdW interface, expanding the horizons of emergent phenomena achievable in vdW heterostructures.

論文

Simulation-based Level 2 multi-unit PRA using RAVEN and a simplified thermal-hydraulic code

Zheng, X.; Mandelli, D.*; Alfonsi, A.*; Smith, C.*; 杉山 智之

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2176 - 2183, 2020/11

The paper introduces a simulation-based Level 2 probabilistic risk assessment (PRA) of a multi-unit nuclear power plant. We propose the methodology by quantifying risk for a station-blackout accident scenario, initialized by a loss-of-offsite-power event. Contrary to classical PRA that applies static models such as event-tree/fault-tree, the analysis is seamlessly integrated with mechanistic simulation and PRA models, including: (1) a simplified thermal-hydraulic code for simulating system behaviors; (2) a Markovian model for the failure mechanism of decay-heat-removal systems, to investigate the interaction between mechanistic simulation and reliability analysis; and (3) classical containment event trees for evaluating containment performances and hydrogen-explosion risk under severe accident conditions. All dynamic and static models, including plant dependencies, are unified within the RAVEN computational framework, applying RAVEN components, External Model, Ensemble Model, and PRA Plugins. The study demonstrates an integrated assessment of risks by considering accident progression and inter-unit system interactions, both time dependent. Statistical data analysis is used to quantifying risk metrics, including core damage frequencies, large early release frequencies and plant damage status. The methodology pertains to modern risk-analysis methodologies such as risk-informed safety margin characterization (RISMC) and dynamic PRA.

論文

Enhancement of the treatment of system interactions in a dynamic PRA tool

田中 洋一; 玉置 等史; Zheng, X.; 杉山 智之

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11

One advantage of dynamic probabilistic risk assessment (PRA) is that it can take into account the timing and ordering of event occurrences based on more explicit simulation of system dynamics. It is expected that dynamic PRA can lead us into a more realistic risk assessment, overcoming some limitations of conventional PRA. Multiple dynamic PRA tools have been developed worldwide, and applied to risk assessment of large industrial facilities such as nuclear power plants and crewed spacecrafts. Japan Atomic Energy Agency has developed the dynamic PRA tool, RAPID (Risk Assessment with Plant Interactive Dynamics), considering the interaction between accident simulation and dysfunctional models of safety-related systems. This paper introduces a recent enhancement of RAPID to treat more complicated simulation interactions from the outside of severe accident codes. It is designed to feed back and forth plant information from simulators to the accident sequence generator. It discusses how the enhancement affects the results of risk assessment, with an example analyzing thermal failure of a safety relief valve in a station blackout accident occurred at a boiling water reactor plant.

論文

Investigation of high-temperature chemical interaction of calcium silicate insulation and cesium hydroxide

Rizaal, M.; 中島 邦久; 斉藤 拓巳*; 逢坂 正彦; 岡本 孝司*

Journal of Nuclear Science and Technology, 57(9), p.1062 - 1073, 2020/09

 被引用回数:3 パーセンタイル:59.68(Nuclear Science & Technology)

福島第一原子力発電所2号機においてペデスタル内よりもペデスタル外で線量が高くなっている現象が見つかっている。この線量の上昇については、原子炉格納容器内の配管に使用されている保温材(ケイ酸カルシウム)がガス状あるいは粒子状となって沈着したセシウム(Cs)と化学反応を起こして固着するとともに破損してペデスタル外に堆積することで線量が上昇した可能性があると考えている。そこで、本研究では、化学反応の有無を調べるため、反応温度等を調べることのできる熱重量示差熱分析装置(TG-DTA)を用いて、水素-水蒸気含有雰囲気下、最高1100$$^{circ}$$Cまで温度を上昇させて、主なセシウム化合物の一つである水酸化セシウムと保温材との混合物に対して分析を行った。その結果、575-730$$^{circ}$$Cの範囲で反応が起こり、試験後試料のX線回折パターンや元素分析機能付き走査型電子顕微鏡(SEM/EDS)による試料表面の元素分布の結果から、保温材の構成物質であるケイ素(Si)に加え、不純物として含まれるアルミニウム(Al)と安定な化合物(CsAlSiO$$_{4}$$)を形成することが分かった。したがって、ペデスタル外で見つかった高線量の原因として、保温材が関係する可能性があることが分かった。

論文

Numerical simulation of liquid jet behavior in shallow pool by interface tracking method

鈴木 貴行*; 吉田 啓之; 堀口 直樹; 山村 聡太*; 阿部 豊*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

In the severe accident (SA) of nuclear reactors, fuel and components melt, and melted materials fall to a lower part of a reactor vessel. In the lower part of a reactor vessel, in some sections of the SAs, it is considered that there is a water pool. Then, the melted core materials fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant may occur. This process is called a fuel-coolant interaction (FCI). FCI is one of the important phenomena to consider the coolability and distribution of core materials. In this study, the numerical simulation of jet breakup phenomena with a shallow pool was performed by using the developed method (TPFIT). We try to understand the hydrodynamic interaction under various, such as penetration, reach to the bottom, spread, accumulation of the molten material jet. Also, we evaluated a detailed jet spread behavior and examined the influence of lattice resolution and the contact angle. Furthermore, the diameters of atomized droplets were evaluated by using numerical simulation data.

論文

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; 三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

 被引用回数:1 パーセンタイル:24.6(Nuclear Science & Technology)

To better understand the failure limit of fuel cladding during the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA), pre-cracked and hydrided cladding samples with base metal final heat-treatment status of cold worked (CW) and recrystallized (RX) were tested under biaxial stress conditions (axial to hoop strain ratios of 0 and 0.5). Displacement-controlled biaxial-expansion-due-to-compression (biaxial-EDC) tests were performed to obtain the hoop strain at failure (failure strain) of the samples. The conversion of the failure strains to J-integral at failure by finite-element analysis involving data of stress-relieved (SR) cladding specimens from our previous study revealed that the failure limit in the dimension of J-integral at failure unifies the effects of pre-crack depth. About 30 to 50 percent reduction in the J-integral at failure was observed as the strain ratio increased from 0 to 0.5 irrespective of the annealing type, pre-crack depth, and hydrogen content. the rate of fractional decreases of J-integral at failure with increase of hydrogen content are in the order of CW$$>$$SR$$>$$RX, which are essentially independent of strain ratio for the CW and SR samples. The results were incorporated into the failure prediction model of the JAEA's fuel performance code in the form of a correction factor that considers the biaxial loading effect.

論文

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

谷口 良徳; 宇田川 豊; 天谷 政樹

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

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

The fuel-failure-limit data obtained in the simulated reactivity-initiated-accident experiment SPERT-CDC 859 (SPERT859) has entailed a lot of discussions if it represents fuel-failure behavior of typical commercial LWRs for its specific pre-irradiation condition and fuel state. The fuel-rod conditions before and during SPERT859 were thus assessed by the fuel-performance codes FEMAXI-8 and RANNS with focusing on cladding corrosion and its effect on the failure limit of the test rod. The analysis showed that the fuel cladding was probably excessively corroded even when the influential calculation conditions such as fuel swelling and creep models were determined so that the lowest limit of the cladding oxide layer thickness was captured. Such assumption of excessive cladding corrosion during pre-irradiation well explains not only the test-rod state before pulse irradiation but also the fuel-failure limit observed. Such understanding undermines anew the representativeness of the test data as a direct basis of safety evaluation for LWR fuels.

論文

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

宇田川 豊; 三原 武; 谷口 良徳; 垣内 一雄; 天谷 政樹

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 被引用回数:1 パーセンタイル:24.6(Nuclear Science & Technology)

This paper reports a computer-code analysis on the base-irradiation behavior of the chromia-and-alumina-doped BWR rod irradiated to 64 GWd/t in Oskarshamn-3, Sweden, and subjected to the reactivity-initiated-accident (RIA) test OS-1, which resulted in a fuel failure due to pellet-cladding mechanical interaction (PCMI) at the lowest fuel-enthalpy increase in all the BWR tests ever performed. The inverse calculation which utilized post-irradiation examination data as its constraint conditions revealed that the OS-1 rod had very likely experienced more intense PCMI loading due to higher swelling rate during base irradiation than other BWR rods subjected to previous RIA tests and thus had been prone to experience enhanced radial-hydride formation. The significant difference in the cladding hoop-stress more than 50 MPa discriminates the OS-1 rod from other BWR rods and supports the interpretation that enhanced radial-hydrides formation differentiated the PCMI-failure behavior observed in the test OS-1 from the previous BWR-fuel tests.

論文

Experimental and analytical investigation of formation and cooling phenomena in high temperature debris bed

堀田 亮年*; 秋葉 美幸*; 森田 彰伸*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk$"o$ld, P.*; et al.

Journal of Nuclear Science and Technology, 57(4), p.353 - 369, 2020/04

 被引用回数:6 パーセンタイル:63.99(Nuclear Science & Technology)

Key phenomena in the cooling states of debris beds under wet cavity conditions were classified into several groups based on the complicated geometry, nonhomogeneous porosity and volumetric heat of debris beds. These configurations may change due to the molten jet breakup, droplet agglomeration, anisotropic melt spreading, two-phase flow in a debris bed, particle self-leveling and penetration of molten metals into a particle bed. The modular code system THERMOS was designed for evaluating the cooling states of underwater debris beds. Three additional tests, DEFOR-A, PULiMS and REMCOD were employed to validate implemented models. This paper summarizes the entire test plan and representative data trends prior to starting individual data analyses and validations of specific models that are planned to be performed in the later phases. It also tries to report research questions to be answered in future works, such as various scales of melt-coolant interactions observed in the PULiMS tests.

論文

Leaching behavior of prototypical Corium samples; A Step to understand the interactions between the fuel debris and water at the Fukushima Daiichi reactors

仲吉 彬; Jegou, C.*; De Windt, L.*; Perrin, S.*; 鷲谷 忠博

Nuclear Engineering and Design, 360, p.110522_1 - 110522_18, 2020/04

 被引用回数:8 パーセンタイル:90.7(Nuclear Science & Technology)

Simulated in-vessel and ex-vessel fuel debris, fabricated in the Colima experimental facility set up in the PLINIUS platform at CEA Cadarache, were selected and leaching experiments were carried out under oxidizing conditions. In parallel, geochemical modeling was performed to better understand the experimental concentrations, pH evolutions and secondary phase's formation. Finally, the Fractional Release Rates of the (U, Zr)O$$_{2}$$ matrix for the two kinds of samples (in-vessel and ex-vessel) were found to be close to or one order of magnitude lower than that of SF under oxidizing conditions (from 10$$^{-6}$$ to 10$$^{-7}$$ per day), but the release processes are different.

論文

Electric field effect on the magnetic domain wall creep velocity in Pt/Co/Pd structures with different Co thicknesses

小山 知弘*; 家田 淳一; 千葉 大地*

Applied Physics Letters, 116(9), p.092405_1 - 092405_5, 2020/03

AA2019-0550.pdf:0.91MB

 被引用回数:3 パーセンタイル:27.08(Physics, Applied)

The electric field (EF) modulation of magnetic domain wall (DW) creep velocity $$v$$ in Pt/Co/Pd structure with perpendicular magnetic anisotropy (MA) has been studied. The structures with different Co thickness $$t_mathrm{Co}$$ up to $$sim 1$$ nm are investigated. In all samples, applying a gate voltage induces a clear change in $$v$$. Thicker samples provide a higher $$v$$ modulation efficiency, and the $$v$$ modulation magnitude of more than a factor of 100 times is observed in the thickest $$t_mathrm{Co}$$ of 0.98 nm. The parameter characterizing the creep motion is significantly affected by the EF, resulting in the modulation of $$v$$. Unlike the $$v$$ case, the MA modulation efficiency decreases with increasing $$t_mathrm{Co}$$. The present results are discussed based on the EF-induced change in the interfacial Dzyaloshinskii-Moriya interaction (iDMI), which has been recently demonstrated in the same structure, and $$t_mathrm{Co}$$ dependence of the DW energy. The $$t_mathrm{Co}$$ dependence of the $$v$$ modulation suggests that the EF effect on the iDMI is more important than the MA.

論文

Electric field control of magnetic domain wall motion via modulation of the Dzyaloshinskii-Moriya interaction

小山 知弘*; 仲谷 栄伸*; 家田 淳一; 千葉 大地*

Science Advances (Internet), 4(12), p.eaav0265_1 - eaav0265_5, 2018/12

AA2018-0306.pdf:0.57MB

 被引用回数:34 パーセンタイル:87.68(Multidisciplinary Sciences)

Pt/Co/Pd非対称構造において電場が磁壁速度を制御できることを示す。ゲート電圧を印加すると、50m/sまでの磁壁速度の著しい変化が観察され、これは以前の研究で観察されたものよりはるかに大きい。さらに、100m/sを超える磁壁速度の明確な変調も確認した。電場による界面のDzyaloshinskii-Moriya相互作用(DMI)の数パーセントまでの変化が、速度変調の原因であることがわかった。ここに示されているDMIを介した速度変化は、電場による異方性変調によって引き起こされるものとは根本的に異なるメカニズムである。本結果は、スピントロニクスデバイスの性能を向上させることができるDMI制御によるスピン構造とダイナミクスの電気的操作への道を開くものである。

論文

OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Impact of number of radial pellet cracks and pellet-clad friction coefficient

Dost$'a$l, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; 天谷 政樹; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

The benchmark on PCMI was initiated by OECD/NEA Expert Group on Reactor Fuel Performance (EGRFP) in June 2015 and is currently in the latter stages of compiling results and preparing the final report. The aim of the benchmark is to improve understanding and modelling of PCMI amongst NEA member organisations. This is being achieved by comparing PCMI predictions of different fuel performance codes for a number of cases. Two of these cases are hypothetical cases aiming to facilitate understanding of the effects of code-to-code differences in fuel performance models. The two remaining cases are actual irradiations, where code predictions are compared with measured data. During analysis of participants' results of the hypothetical cases, the assumptions for number of radial pellet cracks and the pellet-clad friction coefficient (which can be zero, finite or infinite) were identified to be important factors in explaining differences between predictions once pellet-cladding contact occurs. However, these parameters varied in the models and codes used originally by the participants. This fact led to the extension of the benchmark by inclusion of two additional cases, where the number of radial pellet cracks and three different values of the friction coefficient were prescribed in the case definition. Seven calculations from six organisations contributed results were compared and analysed in this paper.

論文

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; 佐藤 一憲; 山路 哲史*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

溶融コリウム・コンクリート相互作用(MCCI)は軽水炉の仮想的シビアアクシデント時の後期フェーズにおいて炉容器外で生じる可能性のある重要事象である。本研究では、MPS法を用いてKITによって実施された模擬物質による成層化溶融プールの実験COMET-L3に対する3次元解析を行った。コリウム/クラスト/コンクリート間の伝熱は粒子間の熱伝導モデルで模擬した。さらに、ケイ酸系コンクリートではケイ酸系析出物の効果によって軸方向と径方向の浸食が異なる可能性が既往研究から示唆されていることから、2つの異なる解析ケースを実施した。解析の結果、MCCIにおいて金属コリウムは酸化物コリウムと全く異なるコンクリート浸食パターンを示しており、アクシデントマネジメントにおける格納系境界の溶融貫通時間の評価に考慮する必要があることが分かった。

論文

Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

上羽 智之; 根本 潤一*; 石谷 行生*; 伊藤 昌弘*

Nuclear Engineering and Design, 331, p.186 - 193, 2018/05

 被引用回数:3 パーセンタイル:40.33(Nuclear Science & Technology)

高速炉MOX燃料ピンの照射挙動を計算するコードと燃料ピン内のCsの挙動に特化して計算するコードを連成することにより、Cs挙動が燃料ピンの熱・機械的挙動に及ぼす影響を解析できるようにした。連成した計算コードを高燃焼度MOX燃料ピンの照射挙動解析に適用し、Csの燃料ピン内軸方向分布やCs化合物による燃料ペレットと被覆管の機械的相互作用などを評価した。

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