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Journal Articles

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07

Journal Articles

Remarks on accepting the 7th Nuclear Fuel Division Award (young investigator award)

Narukawa, Takafumi

Kaku Nenryo, (54-2), P. 3, 2019/07

no abstracts in English

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 122, p.201 - 206, 2018/12

 Percentile:100(Nuclear Science & Technology)

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. This study addresses an improvement of heat-removal capability using heat conduction on the RCCS. As a result, a heat flux removed by the RCCS could be doubled; therefore, it is possible to halve the height of the RCCS or increase the thermal reactor power.

Journal Articles

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.

Journal Articles

Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments

Yumura, Takanori; Amaya, Masaki

Annals of Nuclear Energy, 120, p.798 - 804, 2018/10

 Percentile:100(Nuclear Science & Technology)

Journal Articles

Analyses of LSTF experiment and PWR plant for 5% cold-leg break loss of coolant accident

Watanabe, Tadashi*; Ishigaki, Masahiro*; Katsuyama, Jinya

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10

The analyses of LSTF experiment and PWR plant for 5% cold-leg break LOCA are performed using the RELAP5/MOD3.3 code. The discharge coefficient of critical flow model is determined so as to obtain the agreement of pressure transient between the LSTF experiment and the experimental analysis, and used for the PWR analysis. The characteristics of thermal-hydraulic phenomena in the experiment are shown to be simulated well by the two analyses. The decrease in core differential pressure during the loop-seal clearing is, however, underestimated by the two analyses, and the core heat up is not predicted. The loop flow rates are also underestimated by the two analyses. Although the duration of core heat up during the boil-off period is longer in the experimental analysis, the results of two analyses agree well, and the effect of scaling is found to be small between the experimental analysis and the PWR analysis.

Journal Articles

Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

Journal Articles

Remarks on accepting the 2017 Nuclear Fuel Division Award (presentation award), 1

Narukawa, Takafumi

Kaku Nenryo, (53-2), P. 5, 2018/08

no abstracts in English

Journal Articles

Review of reduction factors by buildings for gamma radiation from radiocaesium deposited on the ground due to fallout

Yoshida-Ouchi, Hiroko*; Matsuda, Norihiro; Saito, Kimiaki

Journal of Environmental Radioactivity, 187, p.32 - 39, 2018/07

 Percentile:100(Environmental Sciences)

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Journal Articles

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

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

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

 Percentile:100(Nuclear Science & Technology)

JAEA Reports

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

Takeda, Takeshi

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

JAEA-Data-Code-2018-003.pdf:3.68MB

Experiment SB-PV-07 was conducted on June 9, 2005 using LSTF. Experiment simulated 1% pressure vessel top small-break LOCA in PWR under total failure of HPI system and nitrogen gas inflow to primary system from ACC tanks. Liquid level in upper-head was found to control break flow rate. Coolant was started to manually inject from HPI system into cold legs as first accident management (AM) action when maximum core exit temperature reached 623 K. Fuel rod surface temperature largely increased because of late and slow response of core exit temperature. SG secondary-side depressurization was initiated by fully opening relief valves as second AM action when primary pressure decreased to 4 MPa. However, second AM action was not effective on primary depressurization until SG secondary-side pressure decreased to primary pressure. Pressure difference became larger between primary and SG secondary sides after ACC tanks started to discharge nitrogen gas.

Journal Articles

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

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

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

 Times Cited Count:2 Percentile:16.17(Materials Science, Multidisciplinary)

Journal Articles

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

Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka

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

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at $$sim$$288$$^{circ}$$C on neutron-irradiated 316L stainless steels (SSs) at $$sim$$12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at $$<$$$$sim$$2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.

Journal Articles

${it In-situ}$ observation of dislocation evolution in ferritic and austenitic stainless steels under tensile deformation by using neutron diffraction

Sato, Shigeo*; Kuroda, Asumi*; Sato, Kozue*; Kumagai, Masayoshi*; Harjo, S.; Tomota, Yo*; Saito, Yoichi*; Todoroki, Hidekazu*; Onuki, Yusuke*; Suzuki, Shigeru*

Tetsu To Hagane, 104(4), p.201 - 207, 2018/00

 Times Cited Count:1 Percentile:52.95(Metallurgy & Metallurgical Engineering)

Journal Articles

RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS

Takeda, Takeshi; Otsu, Iwao

Annals of Nuclear Energy, 109, p.9 - 21, 2017/11

 Times Cited Count:1 Percentile:64.68(Nuclear Science & Technology)

JAEA Reports

Cutting operation of simulated fuel assembly heating examination by AWJ

Abe, Yuta; Nakagiri, Toshio; Watatani, Satoshi*; Maruyama, Shinichiro*

JAEA-Technology 2017-023, 46 Pages, 2017/10

JAEA-Technology-2017-023.pdf:8.01MB

This is a report on Abrasive Water Jet (AWJ) cutting work carried out on specimen, which was used for Simulated Fuel Assembly Heating Examination by Collaborative Laboratories for Advanced Decommissioning Science (CLADS) molten core behavior analysis group in February 2016. The simulated fuel assembly is composed of Zirconia for the outer crucible/simulated fuel, stainless steel for the control blade and Zircaloy (Zr) for the cladding tube/channel box. Therefore, it is necessary to cut at once substances having a wide range of fracture toughness and hardness. Moreover, it is a large specimen with an approximate size of 300 mm. In addition, epoxy resin has high stickiness, making it more difficult to cut. Considering these effects, AWJ cutting was selected. The following two points were devised, and this specimen could be cut with AWJ. If it was not possible to cut at one time like a molten portion of boride, it was repeatedly cut. By using Abrasive Suspension Jet (ASJ) system with higher cutting ability than Abrasive Injection Jet (AIJ, conventional method) system, cutting time was shortened. As a result of this work, the cutting method in Simulated Fuel Assembly Heating Examination was established. Incidentally, in the cutting operation, when the cutting ability was lost at the tip of the AWJ, a curved cut surface, which occurs when the jet flowed away from the feeding direction, could be confirmed at the center of the test body. From the next work, to improve the cutting efficiency, we propose adding a mechanism such as turning the cutting member itself for re-cutting from the exit side of the jet and appropriate traverse speed to protect cut surface.

Journal Articles

Changes of dislocation density and dislocation arrangement during tensile deformation in lath martensitic steels

Harjo, S.; Kawasaki, Takuro; Morooka, Satoshi

Advanced Experimental Mechanics, 2, p.112 - 117, 2017/10

Journal Articles

Evaluation and demonstration of cutting the fuel assembly heating examination by AWJ

Maruyama, Shinichiro*; Watatani, Satoshi*

Mitsui Sumitomo Kensetsu Gijutsu Kenkyu Kaihatsu Hokoku, (15), p.107 - 112, 2017/10

It is essential to estimate characteristics and forms of fuel debris for safe and reliable removing at the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F). For the estimation, melting behavior of fuel assembly in the accident is being researched. To proceed the research, the fuel debris were need to cut, and the abrasive water jet (AWJ) which had enough results for cutting ceramic material or mixed material of zirconium alloy and stainless. The test results demonstrated that AWJ could cut the fuel assembly and accumulated the cutting data which will be subservient when removing the fuel debris in future.

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