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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



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.


Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurised water reactor primary coolant under irradiation, 2; Evaluation of electrochemical corrosion potential

端 邦樹; 塙 悟史; 知見 康弘; 内田 俊介; Lister, D. H.*

Journal of Nuclear Science and Technology, 14 Pages, 2022/00

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



Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Kenzhina, I.*; 石塚 悦男; Ho, H. Q.; 坂本 直樹*; 奥村 啓介; 竹本 紀之; Chikhray, Y.*

Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03

JMTRとJRR-3Mの運転中に一次冷却水へ放出されるトリチウムについて研究してきた結果、ベリリウム中性子反射体の二段核反応による$$^{6}$$Li(n$$_{t}$$,$$alpha$$)$$^{3}$$Hで生成する反跳トリチウムが主要因であることが明らかになった。この結果から、一次冷却水へ放出するトリチウムを少なくするためには、ベリリウム中性子反射体の表面積を小さくするか、他の材料で反跳トリチウムを遮蔽する必要がある。本報告では、ベリリウム中性子反射体のトリチウム反跳防止材の概念検討として、Al, Ti, V, Ni, Zr等の多様な材料を候補材として、障壁厚み、長期運転後の放射能、反応度への影響を評価した。この結果、Alがベリリウム中性子反射体のトリチウム反跳防止材として適した候補材になり得るとの結果を得た。


Evaluation of tritium release into primary coolant for research and testing reactors

Kenzhina, I.*; 石塚 悦男; 奥村 啓介; Ho, H. Q.; 竹本 紀之; Chikhray, Y.*

Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01

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



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.


Feasibility study of tritium recoil barrier for neutron reflectors

石塚 悦男; 坂本 直樹*

Physical Sciences and Technology, 6(2), p.60 - 63, 2019/12



Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

加治 芳行; 根本 義之; 永武 拓; 吉田 啓之; 東條 匡志*; 後藤 大輔*; 西村 聡*; 鈴木 洋明*; 大和 正明*; 渡辺 聡*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05



Calculation of tritium release from driver fuels into primary coolant of research reactors

Ho, H. Q.; 石塚 悦男

Physical Sciences and Technology, 5(2), p.53 - 56, 2019/00




石塚 悦男; Kenzhina, I.*; 奥村 啓介; Ho, H. Q.; 竹本 紀之; Chikhray, Y.*

JAEA-Technology 2018-010, 33 Pages, 2018/11




Behaviors of high-burnup LWR fuels with improved materials under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

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

Fuels for light water reactors (LWRs) which consist of improved cladding materials and pellets have been developed by utilities and fuel vendors to acquire better fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate adequacy of the present regulatory criteria in Japan and safety margins regarding the fuel with improved materials, Japan Atomic Energy Agency (JAEA) has conducted ALPS-II program sponsored by Nuclear Regulation Authority (NRA), Japan. In this program, the tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) have been performed on the high burnup advanced fuels irradiated in commercial PWR or BWR in Europe. This paper presents recent results obtained in this program with respect to RIA, and main results of LOCA experiments, which have been obtained in the ALPS-II program, are summarized.


Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08

 被引用回数:13 パーセンタイル:81.47(Nuclear Science & Technology)

An experiment was conducted for OECD/NEA ROSA-2 Project using LSTF, which simulated 17% hot leg intermediate-break LOCA in PWR. Core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on upper core plate. Results of uncertainty analysis with RELAP5/MOD3.3 code clarified influences of combination of multiple uncertain parameters on peak cladding temperature within defined uncertain ranges. An experiment was performed for OECD/NEA PKL-3 Project with PKL. The LSTF test simulated PWR 1% hot leg small-break LOCA with steam generator secondary-side depressurization as accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for primary pressure, core collapsed liquid level, and cladding surface temperature probably due to effects of differences between LSTF and PKL in configuration, geometry, and volumetric size.


Evaluation of tritium release curve in primary coolant of research reactors

石塚 悦男; Kenzhina, I. E.*

Physical Sciences and Technology, 4(1), p.27 - 33, 2018/06



JASMINE Version 3による溶融燃料-冷却材相互作用SERENA2実験解析

堀田 亮年*; 森田 彰伸*; 梶本 光廣*; 丸山 結

日本原子力学会和文論文誌, 16(3), p.139 - 152, 2017/09

Among twelve FCI cases conducted in the OECD/NEA/CSNI/SERENA2 test series using two facilities, six steam explosion cases, five from TROI and one from KROTOS, were analyzed by JASMINE V.3. Major model parameters were categorized into "focused zone", a core part of interest, and "peripheral zone", the initial and boundary conditions given intentionally for each test case. For the former, base values established through past validation studies of JASMINE V.3 were applied. The code was modified to implement the measured distribution of entrained droplet size acquired in TROI-VISU. For the latter, melt release histories were given as a combination of time tables of jet diameter and release velocity that were estimated based on image data and transit timing data of the melt leading edge. The base values were shown to predict impulse responses of SERENA2 systematically with a reasonable error band. A statistical analysis based on the LHS method was performed. Uncertainty ranges were given based on measurement errors and past validation studies in the JASMINE development. Underlying mechanisms causing apparent differences in the mechanical energy conversion ratio between two facilities were studied from the view point of breakup length and trigger timing.


Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

JAEA has conducted a research program called ALPS-II program for advanced fuels of LWRs. In this program, the tests simulating a RIA and a LOCA have been performed on the high burnup advanced fuels irradiated in European commercial reactors. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by the pulse irradiation tests at the NSRR in JAEA. The information about pellet fragmentation etc. during the pulse irradiations was also obtained from post-test examinations on the test rods after the pulse irradiation tests. As for the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the RFEF in JAEA. The fracture limits under LOCA and post-LOCA conditions etc. of the high-burnup advanced fuel cladding have been investigated, and it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined.


Development of the severe accident evaluation method on second coolant leakages from the PHTS in a loop-type sodium-cooled fast reactor

山田 文昭; 今泉 悠也; 西村 正弘; 深野 義隆; 有川 晃弘*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07



The Effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11

 被引用回数:9 パーセンタイル:68.22(Nuclear Science & Technology)

In order to investigate the effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 (Zry-4) cladding tube, laboratory-scale experiments on non-irradiated Zry-4 cladding tube specimens were performed under transient-heating conditions which simulate loss-of-coolant-accident (LOCA) conditions by using an external heating method, and the data obtained were compared to those from a previous study where an internal heating method was used. The maximum circumferential strains ($$varepsilon$$s) of the cladding tube specimens were firstly divided by the engineering hoop stress ($$sigma$$). The divided maximum circumferential strains, ${it k}$s, of the previous study, which used the internal heating method, were then corrected based on the azimuthal temperature difference (ATD) in the cladding tube specimen. The ${it k}$s for the external heating method which was used in this study agreed fairly well with the corrected ${it k}$s obtained in the previous study which employed the internal heating method in the burst temperature range below $$sim$$1200 K. Also, the area of rupture opening tended to increase with increasing of the value which is defined as $$varepsilon$$ multiplied by $$sigma$$. From the results obtained in this study, it was suggested that $$varepsilon$$ and the size of rupture opening of a cladding tube under LOCA-simulated conditions can be estimated mainly by using $$sigma$$, $$varepsilon$$ and ATD in the cladding tube specimen, irrespective of heating methods.



石塚 悦男; Kenzhina, I. E.*; 奥村 啓介; 竹本 紀之; Chikhray, Y.*

JAEA-Technology 2016-022, 35 Pages, 2016/10


試験研究炉の一次冷却水中へのトリチウム放出機構解明の一環として、ベリリウム炉心構成材からの反跳トリチウム放出率を評価するためPHITSコードを用いた場合の計算方法について検討した。この結果、線源に中性子またはトリトンを用いた場合、両者とも反跳トリチウム放出率は同程となったが、トリトン線源の計算速度が2桁程度速いことが明らかとなった。また、トリトン線源を用いて反跳トリチウム放出率を有効数字2桁の精度で求めるためには、単位体積あたりのヒストリー数が2$$times$$10$$^{4}$$ (cm$$^{-3}$$)程度になるまで計算すれば良いことが明らかとなった。更に、トリトン線源を用いてベリリウム炉心構成材の形状と反跳トリチウム放出率の関係を調べたところ、反跳トリチウム放出率はベリリウムの体積当たりの表面積に対して線形となったが、従来の式を使って求めた値の約半分となった。


A Modelling study on water radiolysis for primary coolant in PWR

向井 悟*; 梅原 隆司*; 塙 悟史; 笠原 茂樹; 西山 裕孝

Proceedings of 20th International Conference on Water Chemistry of Nuclear Reactor Systems (NPC 2016) (USB Flash Drive), 9 Pages, 2016/10



The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

 被引用回数:6 パーセンタイル:52.61(Nuclear Science & Technology)

In order to investigate the effect of oxidation and crystal phase condition on the ballooning and rupture behaviors of cladding tube under simulated loss-of-coolant-accident (LOCA) conditions, laboratory-scale experiments were performed in which internally pressurized non-irradiated Zircaloy-4 (Zry-4) cladding specimens were heated to burst in steam and argon gas conditions. Values of the maximum circumferential strain were normalized by dividing them by engineering hoop stress at the time of rupture. The dependence of the normalized value on burst temperature and the relationship between the normalized value and the length, width and area of rupture opening were evaluated. The correlation between the normalized value and the burst temperature suggested that the fraction of the $$beta$$ phase in Zry-4 cladding specimens affected the strain in the specimens and the oxidation of specimens suppressed the amount of ballooning of the specimens. The relationship between the normalized value and the length, width and area of rupture opening indicated that the length, width and area of rupture opening depended on the crystal phase condition in Zry-4 cladding specimens irrespective of atmosphere in the case of the heating rate of $$sim$$3 K/s.

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