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Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 2

山下 拓哉; 佐藤 一憲; 本多 剛*; 野崎 謙一朗*; 鈴木 博之*; Pellegrini, M.*; 酒井 健*; 溝上 伸也*

Nuclear Technology, 206(10), p.1517 - 1537, 2020/10

 被引用回数:0 パーセンタイル:100(Nuclear Science & Technology)

The estimation and understanding of the state of fuel debris and fission products inside the plant is an essential step in the decommissioning of the TEPCO Fukushima Daiichi Nuclear Power Station (1F). However, the direct observation of the plant interior, which is under a high radiation environment. Therefore, in order to understand the plant interior conditions, the comprehensive analysis and evaluation is necessary, based on various measurement data from the plant, analysis of plant data during the accident progression phase and information obtained from computer simulations for this phase. These evaluations can be used to estimate the conditions of the interior of the reactor pressure vessel (RPV) and the primary containment vessel (PCV). Herein, 1F Unit 2 was addressed as the subject to produce an estimated map of the fuel debris distribution from data obtained about the RPV and PCV based on the comprehensive evaluation of various measurement data and information obtained from the accident progression analysis, which were released to the public in June 2018.



井戸村 泰宏; 小野寺 直幸; 山田 進; 山下 晋; 伊奈 拓也*; 今村 俊幸*

スーパーコンピューティングニュース, 22(5), p.18 - 29, 2020/09



Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

阿部 雄太; 山下 拓哉; 佐藤 一憲; 中桐 俊男; 石見 明洋

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04

The authors are developing an experimental technology for simulating severe accident (SA) conditions using simulate fuel material (ZrO$$_{2}$$) that would contribute, not only to Fukushima Daiichi (1F) decommissioning, but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of accident progression behavior. Nontransfer (NTR) type plasma, which has been in practical use with a large torch capacity as high as 2 MW, has the potential to heat subject materials to very high temperatures without selecting the target to be heated. When simulating 1F with SA code, the target of this core-material-melting and relocation (CMMR) experiment was to confirm that NTR plasma has a sufficient heating performance realizing large temperature gradients ($$>$$ 2000 K/m) expected under 1F conditions. The authors selected NTR-type plasma-heating technology that has the advantage of continuous heating in addition to its high-temperature level. The CMMR-2 experiments were carried out in 2017 applying the improved technology (higher heating power and controlled oxygen concentration). The CMMR-2 experiment adopted a 30-min heating period, wherein the power was increased to a level where a large temperature gradient was expected at the lower part of the core under actual 1F accident conditions. Most of the control blade and channel box migrated from the original position. After heating, the simulated fuel assembly was measured by X-ray computed tomography (CT) technology and by electron probe micro-analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective, in terms of applicability of the NTR-type plasma-heating technology to the SA experimental study, was obtained.



山下 卓哉; 沢田 憲良*

JAEA-Research 2019-010, 227 Pages, 2020/03




The CMMR program; BWR core degradation in the CMMR-4 test

山下 拓哉; 佐藤 一憲

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 13 Pages, 2019/03

福島第一原子力発電所事故(1F)廃止に向けては、炉心物質の最終的な分布とその特性を理解することが重要である。これらの特性は明らかに各ユニットの事故進展に左右される。ただし、BWRの事故進展挙動には大きな不確かさが存在する。MAAP-MELCOR Crosswalkによって明らかにされたこの不確かさは、既存の実験データと知識では解決できない。冷却材がBWR炉心から失われると、その後のシナリオはTMI-2に代表されるものと「連続ドレン型」というシナリオに分けられる。この分岐点の主な不確かさは、2つの疑問点としてまとめられる。(Q1)高温で燃料溶融に近接した損傷炉心のガス透過性はどのようなものか。(Q2)燃料溶融前の高温炉心の下方移動とその構造材加熱への影響はどうか。これらの問題に取り組むために、炉心物質の溶融および再配置に関わるCMMR実験が行われた。CMMR-4試験では、スランピング直前の炉心状態に関する有用な情報が得られた。酸化物燃料が溶融に近接する条件での炉心の巨視的なガス透過性の存在が確認され(A1)、実際の炉で生じる可能性の高い燃料柱崩壊があった場合、最も高温の燃料は炉心の高温部から効率的に低温部に移動できず、炉心燃料最高温度の効果的な制限や、支持構造の著しい加熱が生じないことを示唆している(A2)。



山下 拓哉; 山下 勇人; 永江 勇二

鉄と鋼, 105(1), p.96 - 104, 2019/01

 被引用回数:1 パーセンタイル:70.81(Metallurgy & Metallurgical Engineering)

火力・原子力発電プラントの使用温度条件である550度では、フェライト鋼と溶接材の界面で破断が生じる研究結果が報告されている。本研究では、フェライト鋼への溶接時の入熱量が異なる2種類の異材継手を製作した。溶接には高入熱であるプラズマ溶接および低入熱であるティグ溶接をそれぞれ使用した。改良9Cr-1Mo鋼に形成された熱影響部の組織はプラズマ溶接とティグ溶接とで異なっていた。改良9Cr-1Mo鋼/Alloy 600部を用いて試験片を製作し、550度のクリープ試験を実施した。試験より、ティグ溶接を使用した試験片は界面破断し、プラズマ溶接を使用した試験片は界面破断しない結果が得られた。そのため、熱影響部のひずみ分布計測および有限要素解析を実施し、フェライト鋼に形成される熱影響部の変形挙動に着目し界面破断メカニズムを力学的観点で分析した。各溶接法により製作した異材接手のフェライト鋼の界面近傍に形成する熱影響部の特性の違いにより、界面近傍でのひずみ分布に違いが生じることが分かった。界面破断を回避するためには、フェライト鋼界面近傍にクリープひずみ速度が遅い熱影響部を形成させる必要がある。


Communication avoiding multigrid preconditioned conjugate gradient method for extreme scale multiphase CFD simulations

井戸村 泰宏; 伊奈 拓也*; 山下 晋; 小野寺 直幸; 山田 進; 今村 俊幸*

Proceedings of 9th Workshop on Latest Advances in Scalable Algorithms for Large-Scale Systems (ScalA 2018) (Internet), p.17 - 24, 2018/11

 被引用回数:0 パーセンタイル:100

多相流体CFDコードJUPITERの圧力ポアソン方程式に省通信マルチグリッド前処理付共役勾配(CAMGCG)法を適用し、省通信クリロフ部分空間法と計算性能と収束特性を比較した。JUPITERコードにおいてCAMGCGソルバ問題サイズによらずロバーストな収束特性を有し、通信削減と収束特性向上を両立することから、通信削減のみを実現する省通信クリロフ部分空間法に対する優位性が高い。CAMGCGソルバを$$sim 900$$億自由度の大規模多相流体CFDシミュレーションに適用して反復回数を前処理付CG法の$$sim 1/800$$に削減し、Oakforest-PACSにおける8,000ノードまでの良好な強スケーリングとCG法の$$sim 11.6$$倍の性能向上を達成した。


The CMMR program; BWR core degradation in the CMMR-3 test

山下 拓哉; 佐藤 一憲; 阿部 雄太; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 11 Pages, 2018/10



Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

阿部 雄太; 山下 拓哉; 佐藤 一憲; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Authors are developing an experimental technology to realize experiments simulating Severe Accident (SA) conditions using simulant fuel material (ZrO$$_{2}$$ with slight addition of MgO for stabilization) that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. Based on the results of the prototype test, improvement of plasma heating technology was conducted. The Core Material Melting and Relocation (CMMR)-1/-2 experiments were carried out in 2017 with the large-scale simulated fuel assembly (1 m $$times$$ 0.3 m $$phi$$) applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different resulting basically in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment is selected here from the viewpoint of establishing an experimental technology. The CMMR-2 experiment adopted 30-min heating period, the power was increased up to a level so that a large temperature gradient ($$>$$ 2,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. Most of the control blade and the channel box migrated from the original position. After the heating, the simulated fuel assembly was measured by the X-ray Computed Tomography (CT) technology and by Electron Probe Micro Analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective in terms of applicability of the non-transfer type plasma heating technology to the SA experimental study was obtained.


The CMMR program; BWR core degradation in the CMMR-1 and the CMMR-2 tests

山下 拓哉; 佐藤 一憲; 阿部 雄太; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of 12th International Conference of the Croatian Nuclear Society; Nuclear Option for CO$$_{2}$$ Free Energy Generation (USB Flash Drive), p.109_1 - 109_15, 2018/06




山下 晋; 伊奈 拓也*; 井戸村 泰宏; 吉田 啓之

第31回数値流体力学シンポジウム講演論文集(DVD-ROM), 7 Pages, 2017/12



A Numerical simulation method for molten material behavior in nuclear reactors

山下 晋; 伊奈 拓也; 井戸村 泰宏; 吉田 啓之

Nuclear Engineering and Design, 322, p.301 - 312, 2017/10

 被引用回数:13 パーセンタイル:7.51(Nuclear Science & Technology)



Metallurgical investigations on creep rupture mechanisms of dissimilar welded joints between Gr.91 and 304SS

山下 拓哉; 永江 勇二; 菊地 浩一*; 山本 賢二*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

A dissimilar welded joint was adopted to achieve higher thermal efficiency and economy levels in nuclear and thermal power plants. 2 types of dissimilar welded joints which were different heat input during welding to the ferrite steels were manufactured. The dissimilar welded joints were made of following materials; the modified 9Cr-1Mo steel (Gr. 91) for the ferritic steels, the 304 stainless steel for the austenitic steels and the Inconel 600 for the filler metals, Welding methods for the modified 9Cr-1Mo steel were used Plasma Arc Welding and Gas Tungsten Arc Welding (GTAW), respectively. Creep tests were conducted. Specimens by GTAW failed in base metal part and interface between the modified 9Cr-1Mo steel and Inconel 600. Interface failure mechanisms were analyzed from a perspective of metallurgy which were precipitation and growth of type I carbide and formation of oxide layer.


日本-IAEA合同原子力エネルギーマネジメントスクールの概要; 2016年

山口 美佳; 日高 昭秀; 生田 優子; 村上 健太*; 富田 明*; 広瀬 大也*; 渡邉 正則*; 上田 欽一*; 生井澤 賢*; 小野瀬 貴利*; et al.

JAEA-Review 2017-002, 60 Pages, 2017/03




The Welded joint strength reduction factors of modified 9Cr-1Mo Steel for the advanced loop-type sodium cooled fast reactor

山下 拓哉; 若井 隆純; 鬼澤 高志; 佐藤 健一郎*; 山本 賢二*

Journal of Pressure Vessel Technology, 138(6), p.061407_1 - 061407_6, 2016/12

 被引用回数:0 パーセンタイル:100(Engineering, Mechanical)

Modified 9Cr-1Mo steel (ASME Gr.91) is widely used in fossil power plants. In the advanced loop type sodium cooled fast reactor, modified 9Cr-1Mo steel is going to be adopted as a structural material. In welded joints of enhanced creep-strength ferritic steels including modified 9Cr-1Mo steel, creep strength may markedly degrade, especially in the long-term region. This phenomenon is known as Type-IV damage. Therefore, considering strength degradation due to Type-IV damage is necessary. In this study, we propose a creep strength curve and a welded joint strength-reduction factor (WJSRF). The creep strength curve of welded joints was proposed by employing a second-order polynomial equation with LMP using the stress range partitioning method. WJSRF was proposed on the basis of design creep rupture stress intensities. The resulting allowable stress was conservative compared with that prescribed in the ASME code. In addition, the design of the hot-leg pipe in the advanced loop type sodium cooled fast reactor was reviewed considering WJSRF.


Strength of 316FR joints welded by Type 316FR/16-8-2 filler metals

山下 拓哉; 永江 勇二; 佐藤 健一郎*; 山本 賢二*

Journal of Pressure Vessel Technology, 138(2), p.024501_1 - 024501_7, 2016/04

 被引用回数:0 パーセンタイル:100(Engineering, Mechanical)

316FR stainless steel is a candidate structural material of JSFR. Two types of weld metals are candidates for 316FR welded joints; 316FR weld metal and 16-8-2 weld metal. This study evaluated the need to consider the welded joint strength reduction factors in 316FR welded joints. To this end, the tensile and creep strengths of Type 316FR and Type 16-8-2 weld metals were measured, and the effect of delta-ferrite in weld metals was evaluated in creep-strength tests of 316FR welded joints. In tensile and creep strengths of 316FR joints welded by both metal types, the welded joint strength reduction factors were immaterial. The creep strength of 316FR welded joints was negligibly affected by delta-ferrite levels from 4.1 FN to 7.0 FN. Furthermore, the tensile and creep strengths of 316FR joints welded by two methods (Tungsten Inert Gas Welding and Shielded Metal Arc Welding) were the same.


除染活動支援と放射線に対する理解支援活動; 原子力機構による環境回復の取組,3

山下 卓哉; 板橋 靖

日本原子力学会誌, 57(10), p.656 - 661, 2015/10



Colossal thermomagnetic response in the exotic superconductor URu$$_2$$Si$$_2$$

山下 卓也*; 下山 祐介*; 芳賀 芳範; 松田 達磨*; 山本 悦嗣; 大貫 惇睦; 住吉 浩明*; 藤本 聡*; Levchenko, A.*; 芝内 孝禎*; et al.

Nature Physics, 11(1), p.17 - 20, 2015/01

 被引用回数:32 パーセンタイル:11.24(Physics, Multidisciplinary)

Observation of a colossal Nernst signal is reported. URu$$_2$$Si$$_2$$ is known as a heavy fermion superconductor. The superconductivity coexists with the so-called hidden-order phase. The Nernst coefficient is enhanced by as large as million times over the theoretically expected value within the standard framework of superconducting fluctuations. Moreover, contrary to the conventional wisdom, the enhancement is more significant with the reduction of the impurity scattering rate. This anomalous Nernst effect intimately reflects the highly unusual superconducting state in URu$$_2$$Si$$_2$$.


Material strength evaluation for 60 years design in Japanese sodium fast reactor

永江 勇二; 鬼澤 高志; 高屋 茂; 山下 拓哉

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07

316FR and Mod.9Cr-1Mo steels are candidates as structural materials for Japanese Sodium Fast Reactor, JSFR. The design life and operation temperature of JSFR is 60 years and 823K, respectively. Time-dependent allowable stress is essential. The evaluation of allowable stresses to 500,000 h is a considerable item. Long term strength is evaluated from a viewpoint of microstructural evaluation related to fracture mechanism. In addition, degradation after long term operation at elevated temperature is important. Aging is considered as one of the degradation. The effect of aging on short term property is analyzed. Material strength standard is also necessary for very thick tube sheets of forgings and small diameter thin walled seamless pipes, which are made of Mod.9Cr-1Mo steel in steam generators. This paper summarized currently available data and information on the above items, and shows path forward to the development of material strength standard for 60 years design in JSFR.


福島除染推進活動; 除染活動推進員による国直轄の除染事業への技術支援

押味 一之; 山下 卓哉

保全学, 12(1), p.17 - 21, 2013/04

「平成23年3月11日に発生した東北地方太平洋沖地震に伴う原子力発電所の事故により放出された放射性物質による環境の汚染への対処に関する特別措置法」に基づき、国直轄の除染事業が環境省によって進められている。国直轄の除染事業を支援するため、平成24年2月1日に日本原子力研究開発機構の福島技術本部に総勢20名の「除染活動推進員」を配置し、除染作業現場での技術指導、住民説明会や戸別除染のための同意書取得等の支援活動を開始した。前報(「特集連載 除染活動の紹介(1)」)では専門家活動による自治体協力について報告したが、本報では除染活動推進員による国直轄の除染事業への技術支援のうち除染作業の立会・技術支援について報告する。

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