※ 半角英数字
 年 ~ 
検索結果: 35 件中 1件目~20件目を表示


Initialising ...



Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...



Revaporization behavior of cesium and iodine compounds from their deposits in the steam-boron atmosphere

Rizaal, M.; 三輪 周平; 鈴木 恵理子; 井元 純平; 逢坂 正彦; Gou$"e$llo, M.*

ACS Omega (Internet), 6(48), p.32695 - 32708, 2021/12

This paper presents our investigation on cesium and iodine compounds revaporization from cesium iodide (CsI) deposits on the surface of stainless steel type 304L, which were initiated by boron and/or steam flow. A dedicated basic experimental facility with a thermal gradient tube (TGT) was used for simulating the phenomena. The number of deposits, the formed chemical compounds, and elemental distribution were analyzed from samples located at temperature range 1000-400 K. In the absence of boron in the gas flow, it was found that the initial deposited CsI at 850 K could be directly re-vaporized as CsI vapor/aerosol or reacted with the carrier gas and stainless steel (Cr$$_{2}$$O$$_{2}$$ layer) to form Cs$$_{2}$$CrO$$_{4}$$ on the former deposited surface. The latter mechanism consequently gave a release of gaseous iodine that was accumulated downstream. After introducing boron to the steam flow, a severe revaporization of iodine deposit at 850 K occurred (more than 70% initial deposit). This was found as a result of the formation of two kinds of cesium borates (Cs$$_{2}$$B$$_{4}$$O$$_{7}$$$$cdot$$5H$$_{2}$$O and CsB$$_{5}$$O$$_{8}$$$$cdot$$4H$$_{2}$$O) which contributed to a large release of gaseous iodine that was capable of reaching outlet of TGT ($$<$$ 400 K). In the case of nuclear severe accident, our study have demonstrated that gaseous iodine could be expected to increase in the colder region of a reactor after late release of boron or a subsequent steam flow after refloods of the reactor, thus posing its near-term risk once leaked to the environment.


The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr$'e$, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; 中村 秀夫

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

WGAMA started on Dec. 31st 1999 to assess and strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in NPP and to facilitate international convergence on safety issues and AM analyses and strategies. WGAMA addresses reactor thermal-hydraulics (Thys), in-vessel behavior of degraded cores, containment behavior and protection, and FP release, transport, deposition and retention, for both current and advanced reactors. This paper summarizes such WGAMA contributions in Thys, CFD and severe accidents, which include the Fukushima-Daiichi accident impacts on the WGAMA activities and their substantial outcomes. Around 50 technical reports have become reference in the related fields, which appear in References. Recommendations in these reports include further research, some of which have given rise to the joint projects conducted or underway within the OECD framework. Ongoing WGAMA activities are numerous and a number of them are to be launched in the near future, which are shortly mentioned too.


Development of ex-vessel phenomena analysis model for multi-scenario simulation system, spectra

内堀 昭寛; 青柳 光裕; 高田 孝; 大島 宏之

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



Chemical forms of uranium evaluated by thermodynamic calculation associated with distribution of core materials in the damaged reactor pressure vessel

池内 宏知; 矢野 公彦; 鷲谷 忠博

Journal of Nuclear Science and Technology, 57(6), p.704 - 718, 2020/06

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

福島第一原子力発電所から取り出された燃料デブリへの効果的な処置方策を提案する上では、燃料デブリ中でUがとりうる化学形についての詳細な調査が不可欠である。特に、アクセス性に乏しい圧力容器内に残留する燃料デブリに関する情報が重要である。本研究では、圧力容器内燃料デブリ中、特にマイナー相におけるUの化学形を評価することを目的とし、1F-2号機の事故進展での材料のリロケーション及び環境変化を考慮した熱力学計算を実施した。組成,温度,酸素量といった計算条件は、既存の事故進展解析の結果から設定した。計算の結果、Uの化学形はFeとOの量によって変化し、Feの少ない領域で$$alpha$$-(Zr,U)(O)、Feの多い領域でFe$$_{2}$$(Zr,U) (Laves相)の生成が顕著であった。還元性条件で生成するこれらの金属相中には数パーセントのUが移行しており、燃料デブリの処置において核物質の化学分離を考慮する場合はこれらの相の生成に留意すべきと考えられる。


Creep deformation analysis of a pipe specimen based on creep damage evaluation method

勝山 仁哉; 山口 義仁; Li, Y.

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07



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



Application of Bayesian approaches to nuclear reactor severe accident analysis

Zheng, X.; 玉置 等史; 塩津 弘之; 杉山 智之; 丸山 結

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 11 Pages, 2017/11

Nuclear reactor severe accident simulation involves uncertainties, which may result from incompleteness of modeling of accident scenarios, selection of alternative models and unrealistic setting of parameters during the numerical simulation, etc. Both deterministic and probabilistic methods are required to reach reasonable estimation of risk for severe accidents. Computational codes are widely used for the deterministic accident simulations. Bayesian approaches, including both parametric and nonparametric, are applied to the simulation-based severe accident researches at Japan Atomic Energy Agency (JAEA). In the paper, an overview of these research activities is introduced: (1) Dirichlet process models, a nonparametric Bayesian approach, are applied to source term uncertainty and sensitivity analyses; (2) Gaussian process models are applied to the optimization for operations of severe accident countermeasures; (3) Nonparametric models, include models based on Dirichlet process and K-nearest neighbors algorithm, are built to predict the chemical forms of fission products. Simplified models are integrated into the integral severe accident code, THALES2/KICHE; (4) We have also launched the research of dynamic probabilistic risk assessment (DPRA), and because a great number of accident scenarios will be generated during DPRA, Bayesian approaches would be useful for the boosting of computational efficiency.


Formation and release of molecular iodine in aqueous phase chemistry during severe accident with seawater injection

城戸 健太朗; 端 邦樹; 丸山 結; 西山 裕孝; 星 陽崇*

NEA/CSNI/R(2016)5 (Internet), p.204 - 212, 2016/05

Seawater injection into the degraded core is one of the measures of accident management as it has been performed at Fukushima Daiichi Nuclear Power Plant. The constituents of seawater deeply relates to the iodine chemistry in the water pool of the suppression chamber, which indicates that it is important to assess their effect on the source term in a severe accident. In the present study, by employing a four-component seawater (SW) model we try to simulate the I$$_2$$ molecules yielding in aqueous solution as the function of time, based on several datasets about chemical reaction kinetics and to evaluate its fraction of the initial inventory released from the solution to gas phase. The amount of I$$_2$$ molecule in gas phase was in proportion as the SW mixing ratio. The combination of bromide and hydrogen-carbonate anions considerably contributes to the behavior of the history of producing I$$_2$$ gas. The oxygen molecules solved from air drastically reduced yielding I$$_2$$ gas by catalytically consuming hydroxyl radicals, while the I$$_2$$ gas increased by the carbon dioxide gas contained in air. The effects of SW and carbon dioxide gas are recommended to be considered in the quantitative discussion about I$$_2$$ gas released from aqueous solution.


Application of Bayesian nonparametric models to the uncertainty and sensitivity analysis of source term in a BWR severe accident

Zheng, X.; 伊藤 裕人; 川口 賢司; 玉置 等史; 丸山 結

Reliability Engineering & System Safety, 138, p.253 - 262, 2015/06

 被引用回数:7 パーセンタイル:40.76(Engineering, Industrial)

An important issue for nuclear severe accident is the source tern uncertainty and sensitivity analysis. Generally, thousands of cases are needed to reach a stable result of sensitivity analysis. Based on the limited data obtained by MELCOR analysis, in which the accident at Unit 2 of the Fukushima Daiichi Nuclear Power Plant is used as an example, an approximate stochastic model has been constructed via Bayesian nonparametrics, specifically, the Dirichlet process. The advantage of a nonparametric model is that any deterministic function between explanatory and response variables is not necessary to be determined. The complexity of model will grow automatically as more actual data is observed. The approximate model saves the computational cost and makes it possible to complement thousands of Monte Carlo computation for uncertainty and sensitivity analysis. Probability density functions of uncertainty analysis by MELCOR and the approximate model are obtained and compared. Two densities show great accordance that proves the good predictive ability of the stochastic model. The appropriateness of the approximate model is further validated by the cross-validation through the comparison with actual MELCOR results. Global sensitivity analysis by Sobol' sensitivity index has been performed with the approximate model. Three input parameters are ranked according to their respective influences on the output uncertainty based on first-order and total effect.



森山 清史; 丸山 結*; 中村 秀夫

JAERI-Research 2002-021, 36 Pages, 2002/11




シビアアクシデントの伝熱流動現象における素過程に関する研究; 粒子法を用いた蒸気爆発素過程の数値シミュレーション, 原子力基礎研究 H10-027-5 (委託研究)

越塚 誠一*; 池田 博和*; Liu, J.*; 岡 芳明*

JAERI-Tech 2002-013, 60 Pages, 2002/03




Creep failure of reactor cooling system piping of nuclear power plant under severe accident conditions

茅野 栄一; 丸山 結; 前田 章雄*; 原田 雄平*; 中村 秀夫; 日高 昭秀; 柴崎 博晶*; 湯地 洋子; 工藤 保; 橋本 和一郎*

Proceedings of the 7th International Conference on Creep and Fatigue at Elevated Temperatures (CREEP7), p.107 - 115, 2001/06



High temperature interaction between zircaloy-4 and stainless steel type 304

永瀬 文久; 大友 隆; 上塚 寛

JAERI-Research 2001-009, 21 Pages, 2001/03




Post-test creep analysis of piping failure tests in WIND project

茅野 栄一; 丸山 結; 湯地 洋子; 柴崎 博晶*; 中村 秀夫; 日高 昭秀; 工藤 保; 橋本 和一郎; 前田 章雄*

JAERI-Conf 2000-015, p.303 - 308, 2000/11



Single U-tube Testing and RELAP5 code analysis of PCCS with horizontal heat exchanger

中村 秀夫; 近藤 昌也; 浅香 英明; 安濃田 良成; 田畑 広明*; 小幡 宏幸*

Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.336 - 343, 2000/00

横型熱交換器を用いた静的格納容器冷却系(PCCS)の性能評価に必要なデータを得るため、単一の水平U字管を用いた管内凝縮伝熱の基礎試験を(株)日本原子力発電との研究協力で行った。口径32mm、伝熱長8mの試験体を用いた基準流動条件(圧力: 7気圧、入口蒸気流量: 1%崩壊熱出力相当、非凝縮性ガス分圧: 1%)での試験等から、高いガス分圧($$leq$$20%)でも良好な除熱性能が得られることや圧力損失、排気、排水の各特性でも良好な結果を得て、横型熱交換器のPCCSへの適用性を確認した。RELAP5/MOD3コードを用いた実験後解析では、オリジナルコードが凝縮伝熱を過小評価したため、基礎伝熱試験をもとに選定した凝縮伝熱やガスによる伝熱劣化のモデル群を組み込むとともに、凝縮終了後の伝熱管内ガス停滞挙動を表現するノード法の考案等で、実験を良好に予測するコード改良ができた。今後、計画中の多次元流動に着目した大型モデル試験の解析を行うとともに、実機での各種過渡におけるPCCS挙動を予測・評価する。


A Real-time prediction technique of severe accident progression in containment for emergency response

石神 努; 小林 健介

Journal of Nuclear Science and Technology, 35(6), p.443 - 453, 1998/06

 被引用回数:1 パーセンタイル:15.58(Nuclear Science & Technology)



Proceedings of the OECD/CSNI Specialists Meeting on Fuel-coolant Interactions, May 19$$sim$$21, 1997, Tokai-mura, Japan

秋山 守*; 山野 憲洋; 杉本 純

JAERI-Conf 97-011, 829 Pages, 1998/01




High temperature interaction between zircaloy-4 and inconel-718

上塚 寛; 永瀬 文久; 大友 隆

Journal of Nuclear Materials, 246(2-3), p.180 - 188, 1997/00

 被引用回数:11 パーセンタイル:66.72(Materials Science, Multidisciplinary)




永瀬 文久; 上塚 寛; 大友 隆

JAERI-Research 95-085, 48 Pages, 1995/11




Analysis of reactor piping behavior against thermal loads during severe accident

橋本 和一郎; 中村 尚彦; 五十嵐 実*; 丸山 結; 杉本 純

The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering, Vol. 3, 0, p.1241 - 1246, 1995/00


35 件中 1件目~20件目を表示