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論文

New insight on the thermal impact on cementitious materials due to high-temperature with water supply; Continuous expansive spalling in water

三浦 泰人*; 宮本 慎太郎*; 丸山 一平*; Aili, A.*; 佐藤 拓未; 永江 勇二; 五十嵐 豪*

Case Studies in Construction Materials, 21, p.e03571_1 - e03571_14, 2024/12

 被引用回数:0 パーセンタイル:0.00(Construction & Building Technology)

In this study, the expansion behavior of cement materials after high-temperature heating and water immersion was observed experimentally. Two experiments were conducted using mortar specimens with different sand-to-cement ratios subjected to different high-temperature histories up to 1000$$^{circ}$$C. In Case 1, the specimens were immersed in water after high-temperature heating and then cooled naturally; in Case 2, the specimens were immersed in water at high temperatures without the cooling process. Based on the results, it was confirmed that lime expansion due to the rehydration of CaO by heating occurred in Case 1. In contrast, dynamic continuous explosive spalling occurred in Case 2 because of water penetration into the specimen at a high temperature. The explosive spalling in water observed in Case 2 is a phenomenon that has not been reported to date. Possible failure mechanisms for lime expansion and continuous expansive spalling in water are suggested.

論文

Development of 3D view application debrisEye for decommissioning of Fukushima Daiichi Nuclear Power Plant

山下 拓哉; 下村 健太; 永江 勇二; 永井 英一*; 安松 智博*; 中島 悟*; 荻野 翔矢*; 溝上 伸也*

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 11 Pages, 2024/05

Internal investigations of the Fukushima Daiichi Nuclear Power Plant (1F) have been conducted, and the internal situation is gradually becoming clearer. In addition, trial debris removal has been conducted and much information is being obtained. The information obtained from the trial debris removal is managed in the decommissioning fundamental research database (debrisWiki), which was established by JAEA and TEPCO. However, it is difficult to understand the entire accident progress only from individual data. Therefore, we developed a 3D view application (debrisEye) for 1F decommissioning. debrisEye was created by Unity. For the CG displayed in debrisEye, pre- and post-accident conditions were constructed. The pre-accident status was created using design information and point cloud data from periodic inspections. The post-accident status was created mainly from the results of the internal investigation. For areas where internal investigations have not yet been obtained, the information in the estimation diagram was reflected. CG displayed on debrisEye can be viewed from any viewpoint and angle using the functionality contained in debrisEye. It is also possible to clipping at any cross section and to show or hide each part. debrisEye can be linked to and used with debrisWiki to write information in any location, thus displaying the analysis results and location of the debris collected. Visual linking of debris analysis results with on-site information is expected to facilitate understanding of accident progress and improve efficiency of decommissioning work.

論文

Numerical analysis of melt penetration behavior in the control rod drive housing of Fukushima Daiichi Nuclear Power Station Unit-2

Li, X.; 山路 哲史*; 佐藤 一憲*; 山下 拓哉; 永江 勇二

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05

For Fukushima Daiichi Nuclear Power Station (1F) Unit-2, the muon radiography investigation results indicate that the fuel debris are largely retained inside the RPV. The current study focuses on the analysis of metallic melt penetration behavior in the CRD housing with Moving Particle Semi-implicit (MPS) method. A three-dimensional CRD housing model with simplified inner structures was established. The injection of SS-Zircaloy eutectic melt into the CRD housing was simulated and its downstream penetration and freezing behavior under vertically varying temperature boundary conditions was analyzed. It is found that the melt would start to freeze and form channel blockages soon after it enters the region with a relatively cold boundary in the downstream.

論文

Formulation of material property formula for calculation of damage in reactor pressure vessel during accident evaluation

下村 健太; 山下 拓哉; 永江 勇二

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05

From the results of the internal investigation of Fukushima Daiichi Nuclear Power Station Unit 2, it was confirmed that part of the fuel assembly (upper tie plate) had fallen to the bottom of the pedestal periphery. From this result, it could be presumed that RPV has a hole large enough for the upper tie plate to drop. However, internal investigations have not revealed where the holes are located at the bottom of the RPV. One of failure mode of the RPV lower head would be assumed to be mechanical failure. In this failure, it is assumed that the RPV lower head will be damaged due to the accumulation of creep damage caused by core material above the creep temperature of the RPV substructure materials falling into the lower plenum. Such damage evaluation is performed by thermohydraulic-structure coupled analysis. In the analysis during accident, the RPV lower head is exposed to high temperature conditions. Therefore, the material properties of the RPV material in the high temperature range are required for evaluation by analysis. In this study, we obtained the strength data of RPV material form the creep temperature range to near the melting point and formulated the material property formulas (elastoplastic stress-strain, creep strain, creep rupture) necessary for mechanical failure evaluation.

論文

The Results of the CLADS-MADE-03 BWR bundle degradation test in steam under 1F Unit 1 postulated conditions

Pshenichnikov, A.; 永江 勇二

Nuclear Engineering and Design, 415, p.112729_1 - 112729_16, 2023/12

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

The paper presents the results of a CLADS-MADE-03 BWR mock-up assembly degradation in high-temperature steam under a transient heating rate of 1 K/s, which are the assumed accident conditions for the beginning of the accident at the Unit 1 of the Fukushima Dai-Ichi Nuclear Power Station. In situ video investigations of the control blade provided visual evidences for the features of the control blade melt relocation. Based on the results, factors like an increase in the amount of core liquefied materials due to change of the melt composition, subsequent change in melt relocation features because of liquid/solid material ratio within the melt and B release and transport by aerosol formation were specially emphasized.

報告書

コールドクルーシブル誘導加熱法を用いた炉心酸化物溶融物中の成分偏析に関する研究(共同研究)

須藤 彩子; M$'e$sz$'a$ros, B.*; 佐藤 拓未; 永江 勇二

JAEA-Research 2023-007, 31 Pages, 2023/11

JAEA-Research-2023-007.pdf:3.61MB

東京電力ホールディングス(株)福島第一原子力発電所で形成された燃料デブリの臨界評価のためには、燃料デブリ中に含まれる成分の偏析傾向を把握することは非常に重要である。特に、燃料デブリ中で中性子吸収材としての役割を担うと考えられるFeおよびGdの分布状況は臨界性に大きな影響を与えると考えられる。本研究では炉心酸化物溶融物中の凝固過程におけるFeおよびGdの偏析傾向を解明するため、コールドクルーシブル誘導加熱法を用い炉心構成材料(UO$$_{2}$$、ZrO$$_{2}$$、FeO、Gd$$_{2}$$O$$_{3}$$、模擬核分裂生成物(MoO$$_{3}$$、Nd$$_{2}$$O$$_{3}$$、SrO、RuO$$_{2}$$))、コンクリート主成分(SiO$$_{2}$$、Al$$_{2}$$O$$_{3}$$、CaO)の溶融凝固試験を行った。本試験では、加熱中溶融試料を徐々に下部に引き抜くことによって、下部から上部に向かって凝固させることを実現した。元素分析の結果、Feは試験体中心付近で試験体下部の最大3.4倍濃縮することがわかった。FeOの初期組成、冷却速度、相分離の有無にかかわらず、すべての試験体でFeの試験体中心部付近への偏析が確認された。このことから、FeOは溶融物中で最終凝固領域に向けて偏析することが考えられる。一方、Gdは試験体中の試験体下部で試験体中心付近の最大2.6倍濃縮した。Gd$$_{2}$$O$$_{3}$$は初期組成1at.%以上の場合、冷却速度、相分離の有無にかかわらず、すべての試験体で試験体下部への偏析が確認された。このことから、Gd$$_{2}$$O$$_{3}$$は溶融物中に1at.%以上含まれる場合、初期に凝固する領域に偏析することが考えられる。一方、模擬核分裂生成物の顕著な偏析は確認されなかった。

論文

Axial variations of oxide layer growth and hydrogen uptake of BWR fuel claddings under steam starvation conditions

坂本 寛*; Adachi, Mika*; 徳島 二之*; 青見 雅樹*; 柴田 裕樹; 永江 勇二; 倉田 正輝

Zirconium in the Nuclear Industry; 20th International Symposium (ASTM STP 1645), p.411 - 432, 2023/11

Steam oxidation tests under steam-starved and non-steam-starved conditions were conducted up to 1573 K using prototypic BWR fuel assembly (four fuel pins and fuel channel box) with approximately 750 mm length. Significant suppression of oxide layer growth and enhancement of hydrogen uptake were found at the downstream positions under the steam-starved conditions. To understand the results obtained in the tests using the prototypic BWR fuel assembly, three separate-effects tests were conducted to obtain a fundamental understanding of mechanism of oxygen and hydrogen uptake and its axial variations and evaluation of hydrogen solubility in oxygen-dissolved Zircaloy-2. It is retrieved that the fuel channel box contributes to the axial variations of oxide layer growth and hydrogen uptake of the fuel pins by acting as a source of hydrogen and a sink of oxygen. The evaluation of hydrogen uptake and release requires a detailed estimation of steam oxidation with time at each elevation.

論文

令和4年度開始「廃炉・汚染水・処理水対策事業費補助金(燃料デブリの性状把握のための分析・推定技術の開発(原子炉圧力容器の損傷状況等の推定のための技術開発)」2022年度最終報告

山下 拓哉; 下村 健太; 永江 勇二; 山路 哲史*; 溝上 伸也; 三次 岳志; 小山 真一

廃炉・汚染水・処理水対策事業事務局ホームページ(インターネット), 53 Pages, 2023/10

令和4年度に原子力機構が補助事業者となって実施した「廃炉・汚染水・処理水対策事業費補助金(燃料デブリの性状把握のための分析・推定技術の開発(原子炉圧力容器の損傷状況等の推定のための技術開発))の成果概要を、最終報告として取りまとめた。本報告資料は、廃炉・汚染水・処理水対策事業費事務局ウェブサイトにて公開される。

論文

Numerical simulation method using a Cartesian grid for oxidation of core materials under steam-starved conditions

山下 晋; 佐藤 拓未; 永江 勇二; 倉田 正輝; 吉田 啓之

Journal of Nuclear Science and Technology, 60(9), p.1029 - 1045, 2023/09

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

We newly developed a detailed simulation method for the oxide layer growth/recession under steam-starved conditions using computational fluid dynamics (CFD) methodologies to elaborate the understanding of failure conditions of fuel assemblies during severe accidents. The new method uses the concept of the distance function in a Cartesian grid and is implemented in the original multiphase/multicomponent CFD code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). A distance calculation of the normal direction from the interface is generally difficult in a Cartesian grid. However, the distance function can give a distance normal to the surface of materials by referring to the value of the function. Thus, the growth/recession calculations, which require the distance normal to the interface, become very easy. We checked the availability of JUPITER, considering these models against the verification and validation problems. As a result, we confirmed that JUPITER gives good results, which may contribute to understanding the progress of core degradation under steam-starved conditions.

論文

Study on chemical interaction between UO$$_{2}$$ and Zr at precisely controlled high temperatures

白数 訓子; 佐藤 拓未; 鈴木 晶大*; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06

 被引用回数:2 パーセンタイル:51.90(Nuclear Science & Technology)

ジルカロイ被覆管とUO$$_{2}$$燃料の溶融反応のメカニズム解明に資するため、温度誤差が可能な限り最小となるよう検討を行い、1840$$^{circ}$$Cから2000$$^{circ}$$Cの範囲でZrとUO$$_{2}$$の高温反応試験を実施した。UO$$_{2}$$るつぼにZr試料を装荷し、アルゴン雰囲気中加熱を行い、生成した反応相の成長状況や溶融状態、組織変化の観察を行った。1890 $$^{circ}$$Cから1930 $$^{circ}$$Cで加熱した試料は、丸く変形しており、$$alpha$$-Zr(O)相と、少量のU-Zr-O溶体相で形成されていた。1940$$^{circ}$$C以上で加熱した試料は大きく変形し、急激に溶体形成反応が進行する様子が観測された。U-Zr-O溶体相の形成反応はZr(O)中の酸素濃度に依存し、酸素濃度の低いZr(O)へ反応はどんどん進展する。そして酸素含有量が高いZr(O)中では、U-Zr-O溶体相の生成が抑制されることが確認された。

報告書

事故時の原子炉圧力容器及び炉内構造物の解析評価に用いる強度特性データ集

下村 健太; 山下 拓哉; 永江 勇二

JAEA-Data/Code 2022-012, 270 Pages, 2023/03

JAEA-Data-Code-2022-012.pdf:38.25MB

発電用原子炉である軽水炉において、東京電力ホールディングス株式会社福島第一原子力発電所と同様な全交流電源喪失が発生した場合には、原子炉圧力容器(RPV: Reactor Pressure Vessel)内の冷却機能の喪失、炉内の水位低下による燃料棒の露出、炉心溶融に伴うRPVの破損やRPV破損に伴う炉内の放射線物質の漏えいが発生することが考えられる。事故進展におけるRPVの損傷、溶融した燃料デブリの流出・拡大等の過程を検証、推定することは、廃炉作業を進める上で重要な情報となる。RPVの破損要因については、RPV下部構造部に加えられる荷重・拘束に起因する破損(力学的破損)、低融点金属や高融点酸化物とRPV底部の構造部材との共晶現象による破損(材料間反応による破損)、RPV底部の構造部材の融点近傍での破損が考えられる。破損要因の内、力学的破損については、数値解析(熱流動解析及び構造解析)により検証を行う。このような数値解析を実施する際には、RPV及び炉内構造物を構成する材料(ジルコニウム、炭化ホウ素、ステンレス鋼、ニッケル合金、低合金鋼等)の融点近傍までの伝熱特性(熱伝導率、比熱、密度)や材料特性(熱膨張係数、ヤング率、ポアソン比、引張、クリープ)が必要となる。本資料においては、公開文献を基に、RPV及び炉内構造物を構成する各材料の融点近傍までの母材の特性データをデータ集として取りまとめた。なお、RPV及び炉内構造物を構成する構造物の中には溶接部も存在するため、今回限られたデータであるが、溶接部に関する特性データも併せて示した。

論文

Thermodynamic analysis for solidification path of simulated ex-vessel corium

佐藤 拓未; 永江 勇二; 倉田 正輝; Quaini, A.*; Gu$'e$neau, C.*

CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 79, p.102481_1 - 102481_11, 2022/12

 被引用回数:1 パーセンタイル:8.33(Thermodynamics)

Investigation of the primary containment vessel inside the Fukushima Daiichi Nuclear Power Station showed that a significant amount of the molten corium reached the bottom of the pedestal region. The molten corium and concrete likely caused a complex interaction called Molten Corium Concrete Interaction. The solidification hysteresis of these ex-vessel debris significantly influences its properties. We performed a thermodynamic analysis using the TAF-ID database to infer the solidification path of U-Zr-Al-Ca-Si-O molten corium, which was chosen for a prototypic system of ex-vessel debris. The solidification path for the CaO-rich sim-corium showed that (i) melting as a single liquid phase above 2430 K, (ii) selective solidification of the oxide-rich corium mainly consisted of fuel materials, and (iii) solidification of the remaining materials as a silicate matrix. In contrast, the solidification path for the SiO$$_{2}$$-rich corium indicated that (i) formation of liquid miscibility gap above 2200 K between U-rich and Zr-rich oxidic melts, (ii) individual precipitation of solid phases in each liquid phase.

論文

X ray spectroscopy on $$Xi$$$$^-$$ atoms (J-PARC E03, E07 and future)

山本 剛史; 藤田 真奈美; 後神 利志*; 原田 健志*; 早川 修平*; 細見 健二; 市川 裕大; 石川 勇二*; 鎌田 健人*; 叶内 萌香*; et al.

EPJ Web of Conferences, 271, p.03001_1 - 03001_5, 2022/11

X-ray spectroscopy of hadronic atoms give us various information on the strong interaction between hadrons and nuclei. At J-PARC, we are aiming for the world-first detection of the X rays from atoms with a doubly strange hyperon, $$Xi$$$$^-$$. Recently, two experiments, J-PARC E07 and J-PARC E03, have been performed for the detection of X rays from $$Xi$$$$^-$$ atoms. Overview and status of these measurements will be presented. We will also discuss future plan of X-ray spectroscopy on $$Xi$$$$^-$$-atoms, which can be performed together with high resolution $$Xi$$$$^-$$ hyper nuclear spectroscopy using newly constructed S-2S spectrometer. Preparation status will be shown in this contribution.

論文

Thermodynamic evaluation on solidification path for U-Zr-Fe-O corium

多木 寛; 永江 勇二; 倉田 正輝

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10

The analysis of small samples retrieved from the inside of Primary Containment Vessel (PCV) of Units 1, 2, and 3 at Fukushima Daiichi Nuclear Power Station detected various types of U-bearing particles. Elucidation of the formation mechanism of these particles is expected to be good practice for real debris characterization. In this study, we attempted to analyze the solidification path of U-bearing particles by the thermodynamic approach. From the thermodynamic solidification path patterns analysis, pattern II (at low oxidation condition) was identified as the available one for Units 1 & 2 particles, whereas pattern IV (at high oxidation condition) would be additionally possibly for Unit 3. From these thermodynamic analyzes, the following characteristic are speculated for the debris in PCV: 1) The debris are likely to have been solidified by gradual cooling from high temperatures to intermediate temperatures and solid-solid transition at lower temperatures may be limited. 2) Units 1 & 2 debris might be exposed to slightly hypo-stoichiometric conditions than Unit 3, and whereas Unit 3 debris might have a wider variation in the oxidation degree.

論文

An Investigation of the microstructure and phase composition of the Zr bearing metallic debris in a bypass channel of a BWR fuel after the exothermic reaction in the CLADS-MADE-04 test

Pshenichnikov, A.; 倉田 正輝; 永江 勇二

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10

CLADS-MADE-04は、下部コア領域での溶融伝播挙動の理解を目的としたシリーズの次のテストである。この寄稿では、電子プローブ マイクロアナライザー(EPMA)によって調査された金属破片の微細構造を含む試験後分析の最近の結果について説明する。テスト中、制御棒ブレードの溶融は、比較的ゆっくりと(数cm/分)急激な強い熱放出の波で発生し、最も高温の領域から、劣化しつつある制御棒ブレードとチャネルボックスに沿って下方に広がり、ジルカロイ-4で作られた壁を消費した。サンプル支持板にも大きな損傷が発生した。このような金属リッチ破片の微細構造の調査により、強化された局所コア劣化のメカニズムを理解できるようになる。EPMAによる相同定を徹底した上で、放熱性が高く周囲への拡散の可能性があることを確認する必要がある。Fe-B共晶デブリとZr-Fe共晶デブリの違いについて概説する。これは、下部炉心プレートのメルトスルーと、下部プレナムへのZr-Fe溶融材料の進行の可能性を理解するために特に重要である。

論文

シビアアクシデント解析・実験の最新技術動向,2; 福島第一原子力発電所における原子炉圧力容器破損メカニズムの解明に向けた取り組み

間所 寛*; 永江 勇二

日本原子力学会誌ATOMO$$Sigma$$, 64(9), p.500 - 503, 2022/09

原子炉過酷事故解析などによる福島第一原子力発電所の事故進展解析では、特に原子炉圧力容器(RPV)下部ヘッドの破損から燃料デブリのペデスタル底部への流出に関して不確かさが大きく、モデル高度化が喫緊の課題となっている。また、2号機の内部調査の結果、ペデスタル下部構造物に目立った損傷が見られないことから、RPV下部ヘッドから流出した燃料デブリは溶融金属主体で多くの燃料酸化物は未溶融であった可能性が指摘されている。そこで、廃炉環境国際共同研究センターでは、固液混合状態での圧力容器破損挙動の理解に取り組んできた。本報では、(1)下部ヘッド溶融プールの熱的挙動を把握するための試験(LIVE試験)と、(2)溶融金属による圧力容器下部ヘッド破損挙動を把握するための試験(ELSA試験)の概要を報告する。

論文

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

AA2022-0018.pdf:8.64MB

 被引用回数:5 パーセンタイル:45.58(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.

論文

Ten years of Fukushima Dai-ichi post-accident research on the degradation phenomenology of the BWR core components

Pshenichnikov, A.; 柴田 裕樹; 山下 拓哉; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 59(3), p.267 - 291, 2022/03

 被引用回数:3 パーセンタイル:33.11(Nuclear Science & Technology)

The paper reviews the results of the JAEA and some International activities over the last ten years of research on the understanding of the core components melting and debris formation in boiling water reactors.

論文

Raman investigation of the CLADS-MADE-02 test debris to confirm the mechanism of the volatile and non-volatile boron compounds formation

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Proceedings of TopFuel 2021 (Internet), 12 Pages, 2021/10

The results of the several recent tests performed in JAEA/CLADS are outlined in this paper. However, particular point of this work is focused on the interesting effect that was found on the debris, that contained partially reacted B$$_{4}$$C (control blade debris). A Raman investigation of the control blade metallic debris helped to refine the governing mechanism of the B-compounds formation and transport, which is probably specific mostly for BWRs due to unique bundle configuration and materials morphology. All these factors may directly influence the accident progression in BWR and influence the final debris properties.

論文

A BWR control blade degradation observed in situ during a CLADS-MADE-02 test under Fukushima Dai-Ichi Unit 3 postulated conditions

Pshenichnikov, A.; 倉田 正輝; 永江 勇二

Journal of Nuclear Science and Technology, 58(9), p.1025 - 1037, 2021/09

 被引用回数:5 パーセンタイル:42.67(Nuclear Science & Technology)

The paper summarizes the results of the control blade degradation test CLADS-MADE-02 performed in JAEA. The test focused at the beginning phase of the accident at Fukushima Dai-Ichi (1F) Unit 3. The investigation provided important data, especially on the temperature history, exhaust gas measurement and in situ video of metallic debris formation and relocation to the colder elevations under the test scenario, which reproduced oxidizing conditions during the initial phase of the 1F Unit 3 reactor heat-up. Based on the test results, some decommissioning related conclusions concerning the formation of new B-rich phases containing Cr and Fe were made.

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