※ 半角英数字
 年 ~   年
検索結果: 2770 件中 1件目~20件目を表示


Initialising ...



Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...


Initialising ...



Some characteristics of gas-liquid two-phase flow in vertical large-diameter channels

Shen, X.*; Schlegel, J. P.*; 日引 俊*; 中村 秀夫

Nuclear Engineering and Design, 333, p.87 - 98, 2018/07

Two phase flows in large-diameter channels are important to efficiently and safely transfer mass and energy in a wide variety of applications including nuclear power plants. Two-phase flows in vertical large-diameter channels, however, show much more complex multi-dimensional nature than those in small diameter channels. Various constitutive equations are required to mathematically close the model to predict two-phase flows with two-fluid model. Validations of the constitutive equations require extensive experiment effort. This paper summarizes the recent experimental studies on two-phase flows in vertical large-diameter channels, which includes measuring technique and available databases. Then, a comprehensive review of constitutive equations is provided covering flow regime transition criteria, drift-flux correlations, interfacial area concentration correlations and one- and two-group interfacial area transport equation(s), with discussions on typical characteristics of large-diameter channel flows. Recent 1D numerical simulations of large-diameter channel flows is reviewed too. Finally, future research directions are suggested.


Estimation of radiocesium dietary intake from time series data of radiocesium concentrations in sewer sludge

Pratama, M. A.; 高原 省五; 宗像 雅広; 米田 稔*

Environment International, 115, p.196 - 204, 2018/06

After the Fukushima accident, it became important to determine the quantity of radionuclide ingested by inhabitants. The most common methods currently used to obtain such data are the market basket (MB) and duplicate (DP) methods. The newly proposed method, which we designate as SL, consists of three steps: (1) the separation of wet weather and dry weather data, (2) determining the mass balance of the wastewater treatment plant (WWTP), and (3) developing a reverse biokinetic model to relate the amount of radionuclides ingested to the amounts contained in the sewer sludge. We tested the new method using the time-dependent radiocesium concentrations in sewer sludge from the WWTP in Fukushima City. The results from the SL method agreed to those from the MB while overestimated those from DP method. The trend lines for all three methods, however, are in good agreement. Sensitivity analyses of SL method indicate further studies on uncertainties of sensitive parameters are deemed necessary to improve the accuracy of the method.


Migration behavior of gaseous ruthenium tetroxide under boiling and drying accident condition in reprocessing plant

吉田 尚生; 大野 卓也; 天野 祐希; 阿部 仁

Journal of Nuclear Science and Technology, 55(6), p.599 - 604, 2018/06

再処理施設における高レベル濃縮廃液の蒸発乾固事故の際に、気体状ルテニウム化合物が環境中へ放出される可能性がある。この事故事象の安全評価に資するために、気体状ルテニウム化合物(四酸化ルテニウム)の移行挙動およびLeak Path Factorを実験的に評価した。施設内の移行経路を模擬したルテニウム気相部移行試験装置および事故時の気相組成を模擬した硝酸含有水蒸気を用いて試験を行った。対照実験として、乾燥空気および水蒸気条件を用いた試験を行った。結果として、ルテニウムは乾燥空気および水蒸気雰囲気下では移行経路に沈着した一方、硝酸含有水蒸気雰囲気下では沈着することなく移行経路を通過した。これらの結果は、気体状ルテニウムの移行挙動が気相条件の影響を受けることを示唆している。


Development of stress intensity factors for subsurface flaws in plates subjected to polynomial stress distributions

Lu, K.; 真野 晃宏; 勝山 仁哉; Li, Y.; 岩松 史則*

Journal of Pressure Vessel Technology, 140(3), p.031201_1 - 031201_11, 2018/06

The stress intensity factor (SIF) solutions for subsurface flaws near the free surfaces of components, which are known to be important in engineering applications, have not been provided yet. Thus, in this paper, SIF solutions for subsurface flaws near the free surfaces in flat plates were numerically investigated based on finite element analyses. The flaws with aspect ratios a/l = 0.0, 0.1, 0.2, 0.3, 0.4 and 0.5, the normalized ratios a/d = 0.0, 0.1, 0.2, 0.4, 0.6 and 0.8, and d/t = 0.01 and 0.10 were taken into account, where a is the half flaw depth, l is the flaw length, d is the distance from the center of the subsurface flaw to the nearest free surface and t is the wall thickness. Fourth-order polynomial stress distribution in the thickness direction was considered. In addition, the developed SIF solutions were incorporated into a Japanese probabilistic fracture mechanics (PFM) code, and PFM analyses were performed for a Japanese reactor pressure vessel containing a subsurface flaw near the inner surface. The PFM analysis results indicate that the obtained SIF solutions are effective in engineering applications.


Uncertainty quantification of fracture boundary of pre-hydrided Zircaloy-4 cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Nuclear Engineering and Design, 331, p.147 - 152, 2018/05

To quantify the fracture boundary uncertainty for non-irradiated, pre-hydrided Zircaloy-4 cladding tube specimens under loss-of-coolant accident conditions at a light-water reactor, data from integral thermal shock tests obtained by an earlier study are analyzed statistically and the fracture boundary is estimated in terms of probability, as follows. First, a method is proposed to obtain the specimens' fracture probability curve as a function of equivalent cladding reacted (ECR) and initial hydrogen concentration using Bayesian inference with a generalized linear model. A log-probit model is used, modified to reflect the effect of the initial hydrogen concentration on the fracture boundary and the ECR evaluation uncertainty, and scaled to improve convergence. Second, using the modified log-probit model, it is shown that the boundary representing a 5% fracture probability with 95% confidence for the pre-hydrided cladding tube sample is higher than 15% ECR, for initial hydrogen concentrations of up to 800 wppm.


Influence of chemical speciation in reactor cooling system on pH of suppression pool during BWR severe accident

塩津 弘之; 石川 淳; 杉山 智之; 丸山 結

Journal of Nuclear Science and Technology, 55(4), p.363 - 373, 2018/04

The influences of chemical speciation for Cs-I-Te-Mo-Sn-B-C-O-H system, simulating a state in the reactor cooling system (RCS) of BWR, on pH of the suppression chamber (S/C) water pool were analytically investigated with PHREEQC code. Major conditions were chosen on the basis of the outputs from a BWR severe accident analysis by THALES2 code and chemical thermodynamic analysis with VICTORIA2.0 code. The chemical thermodynamic analysis showed that the chemical speciation of important volatile FPs, Cs and I, was strongly influenced by Mo and B$$_{4}$$C control material. As a consequence, pH of the S/C water pool was predicted to range from approximately 6 to 10, depending on the fraction of volatile FPs transported from the RCS to the S/C water pool and the H$$_{2}$$/H$$_{2}$$O ratio associated with the oxygen potential. It was implied that the formation of volatile I species such as I$$_{2}$$ in the S/C water pool was larger by 3 orders at the lowest pH than that at the highest pH.



知見 康弘; 岩田 景子; 飛田 徹; 大津 拓与; 高見澤 悠; 吉本 賢太郎*; 村上 毅*; 塙 悟史; 西山 裕孝

JAEA-Research 2017-018, 122 Pages, 2018/03


原子炉圧力容器の加圧熱衝撃(Pressurized Thermal Shock: PTS)事象に対する構造健全性評価に与える影響項目の一つである高温予荷重(Warm Pre-stress: WPS)効果は、高温時に予め荷重を受けた場合に、冷却中の荷重減少過程では破壊が生じず、低温での再負荷時の破壊靱性が見かけ上増加する現象である。WPS効果については、主として弾性データによって再負荷時の見かけの破壊靱性を予測するための工学的評価モデルが提案されているが、試験片の寸法効果や表面亀裂に対して必要となる弾塑性評価は考慮されていない。本研究では、実機におけるPTS時の過渡事象を模擬した荷重-温度履歴を与える試験(WPS効果確認試験)を行い、WPS効果に対する試験片寸法や荷重-温度履歴の影響を確認するとともに、工学的評価モデルの検証を行った。再負荷時の見かけの破壊靭性について、予荷重時の塑性の程度が高くなると試験結果は工学的評価モデルによる予測結果を下回る傾向が見られた。比較的高い予荷重条件に対しては、塑性成分等を考慮することにより工学的評価モデルの高精度化が可能となる見通しが得られた。



石崎 梓; 眞田 幸尚; 西澤 幸康*; 普天間 章; 宗像 雅広

JAEA-Research 2017-012, 58 Pages, 2018/03




Data report of ROSA/LSTF experiment SB-SG-10; Recovery actions from multiple steam generator tube rupture accident

竹田 武司

JAEA-Data/Code 2018-004, 64 Pages, 2018/03




Data report of ROSA/LSTF experiment SB-PV-07; 1% Pressure vessel top break LOCA with accident management actions and gas inflow

竹田 武司

JAEA-Data/Code 2018-003, 60 Pages, 2018/03





勝山 仁哉; 眞崎 浩一; 宮本 裕平*; Li, Y.

JAEA-Data/Code 2017-015, 229 Pages, 2018/03





眞田 幸尚; 森 愛理; 岩井 毅行; 瀬口 栄作; 松永 祐樹*; 河端 智樹; 豊田 政幸*; 飛田 晋一朗*; 平賀 翔吾; 佐藤 義治; et al.

JAEA-Technology 2017-035, 69 Pages, 2018/02





眞田 幸尚; 森 愛理; 岩井 毅行; 瀬口 栄作; 松永 祐樹*; 河端 智樹; 豊田 政幸*; 飛田 晋一朗*; 平賀 翔吾; 佐藤 義治; et al.

JAEA-Technology 2017-034, 117 Pages, 2018/02




Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Journal of Nuclear Materials, 499, p.528 - 538, 2018/02


For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best model to estimate the fracture probability. It was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.


Deformation behavior of recrystallized and stress-relieved Zircaloy-4 fuel cladding under biaxial stress conditions

三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 55(2), p.151 - 159, 2018/02

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to the failure of high-burnup fuel rods. Biaxial stress states generated by PCMI in Zircaloy cladding may make the cladding more susceptible to failure. In this study, we investigated the deformation behavior of Zircaloy cladding under biaxial stress conditions based on the concept of contours of equal plastic work. The major axis angles of the initial work contours of recrystallized (RX) and stress-relieved (SR) specimens were investigated and it was found that the shapes of the initial work contours of these kinds of specimens were almost symmetric across the direction where the ratio of axial stress to circumferential stress is 1. The shapes of subsequent work contours tended to change for the RX specimen while be the same as the initial for the SR specimen, as deformation proceeded. It was suggested that the textures and slip systems in the RX and SR specimens affect their initial work contours while the slip system in the RX specimens and the residual strain in the SR specimens influence the subsequent work contours.



吉田 一雄; 玉置 等史; 吉田 尚生; 天野 祐希; 阿部 仁

JAEA-Research 2017-015, 18 Pages, 2018/01




Engineering applications using probabilistic aftershock hazard analyses; Aftershock hazard map and load combination of aftershocks and tsunamis

崔 炳賢; 西田 明美; 糸井 達哉*; 高田 毅士*

Geosciences (Internet), 8(1), p.1_1 - 1_22, 2018/01



Dose-reduction effects of vehicles against gamma radiation in the case of a nuclear accident

高原 省五; 渡邊 正敏*; 廣内 淳; 飯島 正史*; 宗像 雅広

Health Physics, 114(1), p.64 - 72, 2018/01

The aim of this paper is to evaluate the dose-reduction effects of vehicles. To achieve this aim, a model for calculating the dose reduction factor (DRF) was developed based on the actual shape and weight of Japanese vehicles. This factor is defined as the ratio of dose rate inside a vehicle to that outside. In addition to model calculation, we evaluated the DRFs by actual measurements in the areas contaminated by the Fukushima accident. A comparison between the simulated and the measured results revealed that the DRFs obtained using the developed models were in good agreement with the results of actual measurements. Using this model, we also evaluated the DRFs for cloudshine and groundshine in the case of a nuclear accident. The evaluations were performed for four vehicle models whose weights were 800-1930 kg. The DRF for cloudshine with photon energy of 0.4-1.5 MeV was 0.66-0.88, and that for groundshine from $$^{137}$$Cs was 0.64-0.73.


Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident

更田 豊志*; 永瀬 文久

Zirconium in the Nuclear Industry; 18th International Symposium (ASTM STP 1597), p.52 - 92, 2018/01



Evaluation of crack growth rates and microstructures near the crack tip of neutron-irradiated austenitic stainless steels in simulated BWR environment

知見 康弘; 笠原 茂樹; 瀬戸 仁史*; 橘内 裕寿*; 越石 正人*; 西山 裕孝

Proceedings of 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00


2770 件中 1件目~20件目を表示