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Journal Articles

On the degradation progression of a BWR control blade under high-temperature steam-starved conditions

Pshenichnikov, A.; Nagae, Yuji; Kurata, Masaki

Mechanical Engineering Journal (Internet), 7(3), p.19-00503_1 - 19-00503_10, 2020/06

Journal Articles

New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1; Introduction of boiling water reactor mock-up assembly degradation test programme

Pshenichnikov, A.; Kurata, Masaki; Bottomley, D.; Sato, Ikken; Nagae, Yuji; Yamazaki, Saishun

Journal of Nuclear Science and Technology, 57(4), p.370 - 379, 2020/04

 Times Cited Count:3 Percentile:31.27(Nuclear Science & Technology)

Journal Articles

Raman characterization of the simulated control blade debris to understand the boric compounds transformations during severe accidents

Pshenichnikov, A.; Nagae, Yuji; Kurata, Masaki

Mechanical Engineering Journal (Internet), 7(2), p.19-00477_1 - 19-00477_8, 2020/04

Journal Articles

Oxidation kinetics of silicon carbide in steam at temperature range of 1400 to 1800$$^{circ}$$C studied by laser heating

Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Bottomley, D.; Furumoto, Kenichiro*

Journal of Nuclear Materials, 529, p.151939_1 - 151939_8, 2020/02

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

Journal Articles

Oxidation of silicon carbide in steam studied by laser heating

Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Furumoto, Kenichiro*; Sato, Hisaki*; Ishibashi, Ryo*; Yamashita, Shinichiro

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09

Journal Articles

Fundamental study on segregation behavior in U-Zr-Fe-O system during solidification process

Sudo, Ayako; Mizusako, Fumiki*; Hoshino, Kuniyoshi*; Sato, Takumi; Nagae, Yuji; Kurata, Masaki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(3), p.111 - 118, 2019/08

Cooling rate of molten core materials during solidification significantly affects the segregation of major constituents of fuel debris. To understand general tendency of the segregation, liquefaction/solidification tests of simulated corium (UO$$_{2}$$, ZrO$$_{2}$$, FeO, B$$_{4}$$C and sim-FP oxides) were performed. Simulated corium was heated up to 2600$$^{circ}$$C under Ar atmosphere and then cooled down with two different cooling processes; furnace cooling (average cooling rate is approximately 744$$^{circ}$$C/min) and slow cooling (cooling rate in 2600$$^{circ}$$C$$sim$$2300$$^{circ}$$C is 5$$^{circ}$$C/min and in 2300$$^{circ}$$C$$sim$$1120$$^{circ}$$C is approximately 788$$^{circ}$$C/min). Element analysis detected three oxide phases with different composition and one metal phase in both solidified samples. Solubility of FeO in these oxide phases was mostly fixed to be 12$$pm$$5at% in both samples, which is in reasonable accordance with the value estimated from UO$$_{2}$$-ZrO$$_{2}$$-FeO phase diagrams. However, a significant grain-growth of one oxide phase, rich in Zr-oxide, was detected only in the slowly cooled sample. The composition of this particular oxide phase is comparable to the initial average composition. The condensation is considered to be caused by the connection of remaining liquid agglomerates during slow solidification.

Journal Articles

Features of a control blade degradation observed ${it in situ}$ during severe accidents in boiling water reactors

Pshenichnikov, A.; Yamazaki, Saishun; Bottomley, D.; Nagae, Yuji; Kurata, Masaki

Journal of Nuclear Science and Technology, 56(5), p.440 - 453, 2019/05

 Times Cited Count:3 Percentile:23.33(Nuclear Science & Technology)

Journal Articles

The Behaviour of materials in case of solidified absorber melt - oxidized BWR channel box interaction revealed after CLADS-MADE-01 test

Pshenichnikov, A.; Kurata, Masaki; Nagae, Yuji; Yamazaki, Saishun

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

Journal Articles

High-temperature interaction between zirconium and UO$$_2$$

Shirasu, Noriko; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

High temperature interaction tests between UO$$_{2}$$ and Zr were performed at around 2173 K, to make clear the UO$$_{2}$$/ $$alpha$$-Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO$$_{2}$$ crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.

Journal Articles

Validation and verification for the melting and eutectic models in JUPITER code

Chai, P.; Yamashita, Susumu; Nagae, Yuji; Kurata, Masaki

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 14 Pages, 2019/03

In order to obtain a precise understanding of molten material behavior inside RPV and to improve the accuracy of the SA code, a new computational fluid dynamics (CFD) code with multi-phase, multi-physics models, which is called JUPITER, was developed. It optimized the algorithms of the multi-phase calculation. Besides, the chemical reactions are also modeled carefully in the code so that the melting process could be treated precisely. A series of verification and validation studies are conducted, which show good agreement with analytical solutions and previous experiments. The capabilities of the multi-physics models in JUPITER code provide us another useful tool to investigate the molten material behaviors in the relevant severe accident scenario.

Journal Articles

Mechanical investigation on interface failure mechanisms of dissimilar welded joints

Yamashita, Takuya; Yamashita, Hayato; Nagae, Yuji

Tetsu To Hagane, 105(1), p.96 - 104, 2019/01

 Times Cited Count:1 Percentile:60.52(Metallurgy & Metallurgical Engineering)

no abstracts in English

Journal Articles

Steam oxidation of silicon carbide at temperatures above 1600$$^{circ}$$C

Pham, V. H.; Nagae, Yuji; Kurata, Masaki

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 6 Pages, 2018/10

Journal Articles

High temperature oxidation test of simulated BWR fuel bundle in steam-starved conditions

Yamazaki, Saishun; Pshenichnikov, A.; Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Tokushima, Kazuyuki*; Aomi, Masaki*; Sakamoto, Kan*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 8 Pages, 2018/10

It is predictively evaluated that degradation of fuel assembly proceeded in a certain steam-starved condition at the early stage of a SA at 1F unit 2 (BWR). As for PWR fuel assembly, effective steam flow rate was properly indicated by normalizing to a unit of g-H$$_{2}$$O/sec/rod which is used as an important parameter for evaluating fuel degradation progression. Due to the inhomogeneous configuration of BWR fuel assembly, the difference of Zry oxidation and hydrogen uptake between the inside and outside of the channel box cannot be properly evaluated by this normalization. Instead of g-H$$_{2}$$O/sec/rod, proper evaluation unit for BWR configuration is necessary. To accumulate Zry oxidation and hydrogen uptake data for steam-starved conditions, high temperature oxidation tests were performed using a simulated BWR fuel bundle sample. The use of equivalent diameter of the cross section of BWR fuel assembly was proposed for normalization of effective steam flow rate.

Journal Articles

The CMMR program; BWR core degradation in the CMMR-3 test

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

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 11 Pages, 2018/10

no abstracts in English

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

The CMMR program; BWR core degradation in the CMMR-1 and the CMMR-2 tests

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

Proceedings of 12th International Conference of the Croatian Nuclear Society; Nuclear Option for CO$$_{2}$$ Free Energy Generation (USB Flash Drive), p.109_1 - 109_15, 2018/06

no abstracts in English

Journal Articles

Application of nontransfer type plasma heating technology for core-material-relocation tests in boiling water reactor severe accident conditions

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

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020901_1 - 020901_8, 2018/04

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mm (height)). An excellent perspective in terms of applicability of the non-transfer plasma heating to melting high melting-temperature materials such as ZrO$$_{2}$$ has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO$$_{2}$$ fuel Phebus-FPT tests. Furthermore, application of EPMA, SEM/EDX and X-ray CT led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the Severe Accident (SA) experimental study was obtained.

Journal Articles

Metallurgical investigations on creep rupture mechanisms of dissimilar welded joints between Gr.91 and 304SS

Yamashita, Takuya; Nagae, Yuji; Kikuchi, Koichi*; Yamamoto, Kenji*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

Journal Articles

Development of non-transfer type plasma heating technology to address CMR behavior during severe accident with BWR design conditions

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

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Journal Articles

Influence of cyclic softening on high temperature material properties in Mod.9Cr-1Mo steel

Onizawa, Takashi; Nagae, Yuji; Kato, Shoichi; Wakai, Takashi

Zairyo, 66(2), p.122 - 129, 2017/02

The applicability of Modified 9Cr-1Mo steel (ASME Grade 91 steel) as the main structural material in advanced loop-type sodium cooled fast reactor has been explored to enhance the safety, the credibility and the economic competitiveness of fast reactor plants. It is well-known that the steel exhibits cyclic softening behavior. Decrease of tensile and creep strength in softened materials has been already reported by other researchers. This paper discusses the relationship between cyclic softening conditions and high temperature material properties. Grade 91 steel was softened by repeat of plastic strain. The softening behavior could be evaluated by the index of the softening rate. Decrease of tensile and creep strength in softened materials can be evaluated by the softening rate and it depends on the cyclic softening conditions.

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