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Investigation of thermal expansion model for evaluation of core support plate reactivity in ATWS event

素都 益武

Journal of Energy and Power Engineering, 14(8), p.251 - 258, 2020/08



Uncertainty evaluation of anticipated transient without scram plant response in the Monju reactor considering reactivity coefficients within the design range

素都 益武; 羽様 平

Journal of Energy and Power Engineering, 13(11), p.393 - 403, 2019/11



A Study on improvement of RANS analysis for erosion of density stratified layer of multicomponent gas by buoyant jet in a containment vessel

安部 諭; 石垣 将宏; 柴本 泰照; 与能本 泰介

Journal of Energy and Power Engineering, 9(7), p.599 - 607, 2015/07



RELAP5 code study of ROSA/LSTF experiments on PWR safety system using steam generator secondary-side depressurization

竹田 武司; 大貫 晃*; 西 弘昭*

Journal of Energy and Power Engineering, 9(5), p.426 - 442, 2015/05

RELAP5 code analyses were performed on two ROSA/LSTF validation tests for PWR safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG (steam generator) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.


Unplanned shutdown frequency prediction of FBR MONJU using fault tree analysis method

素都 益武

Journal of Energy and Power Engineering, 8(7), p.1286 - 1292, 2014/07



Development of an evaluation methodology for fuel discharge in core disruptive accidents of sodium-cooled fast reactors

神山 健司; 飛田 吉春; 鈴木 徹; 松場 賢一

Journal of Energy and Power Engineering, 8(5), p.785 - 793, 2014/05

The purpose of the present study is to develop a methodology to evaluate fuel discharge through the control-rod guide tube (CRGT) during core-disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs) since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by SIMMER. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that SIMMER with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.


Oxygen potential analysis to evaluate irradiation behavior in MOX and MA-bearing MOX fuels

加藤 正人; 安部 智之

Journal of Energy and Power Engineering, 7(10), p.1865 - 1870, 2013/10

Minor actinide bearing MOX fuels have been developed as sodium cooled fast reactor fuels. Content of Am which is one of the minor actinide elements causes oxygen potential to increase. The effects of the oxygen potential increase on the irradiation behavior were evaluated. Profiles of temperature and O/M ratio in the pellets were evaluated to better understand the irradiation behavior. From these data, local oxygen potential in the radial direction of the pellets was calculated, and was compared with free energy of compounds composed of fission products. Based on this comparison, it was concluded that Cs$$_{2}$$MoO$$_{4}$$ was likely formed at pellet periphery of (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{1.98}$$ and (U$$_{0.66}$$Pu$$_{0.3}$$Am$$_{0.016}$$Np$$_{0.016}$$)O$$_{1.976}$$. The extent of cladding tube inner surface oxidation was predicted by using the calculated oxygen potential.


Plant dynamics evaluation of a MONJU ex-vessel fuel storage system during a station blackout

森 健郎; 素都 益武; 本多 慶; 鈴木 悟志*; 大平 博昭

Journal of Energy and Power Engineering, 7(9), p.1644 - 1655, 2013/09



OECD/NEA ROSA project experiment on steam condensation in PWR horizontal legs during large-break LOCA

竹田 武司; 大津 巌; 中村 秀夫

Journal of Energy and Power Engineering, 7(6), p.1009 - 1022, 2013/06

Separate-effect experiment simulating steam direct-contact condensation on emergency core cooling system (ECCS) water in PWR cold legs during reflood phase of large-break LOCA was conducted in OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). A new test section was furnished in the downstream of the LSTF break unit horizontally attached to the cold leg. Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation of the ECCS water injection. Total steam condensation rate estimated from the difference between steam flow rates at the test section inlet and outlet was in proportion to the simulated ECCS water mass flux until the complete condensation of steam. Clear images of high-speed video camera were successfully obtained on droplet behaviors through the viewer of the test section, especially for annular mist flow.


Corrosion behavior of a titanium alloy in hot nitric acid condensate

竹内 正行; 佐野 雄一; 中島 靖雄; 内山 軍蔵; 野島 康夫*; 藤根 幸雄*

Journal of Energy and Power Engineering, 7(6), p.1090 - 1096, 2013/06



Numerical simulations of upper plenum thermal-hydraulics of Monju reactor vessel using high resolution mesh models

大平 博昭; 本多 慶; 素都 益武

Journal of Energy and Power Engineering, 7(4), p.679 - 688, 2013/04

In order to evaluate the upper plenum thermal-hydraulics of the Monju reactor vessel, we have performed detail calculations under the 40% rated power operational condition using high resolution mesh models by a commercial FVM code, FrontFlow/Red. In this study, we applied a high resolution meshes around the flow holes (FHs) on the inner barrel. We mainly made clear that the thermal-hydraulics did not change largely since the flow rates through the FHs were small enough to the total coolant flow rate but were affected largely in case without FHs on the honeycomb structure.


Fundamental study on a grout penetration model for a HLW repository

藤田 朝雄; 新貝 文昭*; 延藤 遵*

Journal of Energy and Power Engineering, 6(8), p.1191 - 1203, 2012/08

高レベル放射性廃棄物地層処分場で仕様可能なグラウト浸透モデルとして、室内試験結果にGustafson and Stillモデルを適用し、その有効性を確認した。


Experimental studies on penetration of pulverized clay-based grout

藤田 朝雄; 杉田 裕; 戸井田 克*

Journal of Energy and Power Engineering, 5(5), p.419 - 427, 2011/05


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