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Journal Articles

New insight on the thermal impact on cementitious materials due to high-temperature with water supply; Continuous expansive spalling in water

Miura, Taito*; Miyamoto, Shintoro*; Maruyama, Ippei*; Aili, A.*; Sato, Takumi; Nagae, Yuji; Igarashi, Go*

Case Studies in Construction Materials, 21, p.e03571_1 - e03571_14, 2024/12

 Times Cited Count:0 Percentile:0.00(Construction & Building Technology)

Journal Articles

Formulation of material property formula for calculation of damage in reactor pressure vessel during accident evaluation

Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05

Journal Articles

Development of 3D view application debrisEye for decommissioning of Fukushima Daiichi Nuclear Power Plant

Yamashita, Takuya; Shimomura, Kenta; Nagae, Yuji; Nagai, Eiichi*; Yasumatsu, Tomohiro*; Nakashima, Satoru*; Ogino, Shoya*; Mizokami, Shinya*

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 11 Pages, 2024/05

Journal Articles

The Results of the CLADS-MADE-03 BWR bundle degradation test in steam under 1F Unit 1 postulated conditions

Pshenichnikov, A.; Nagae, Yuji

Nuclear Engineering and Design, 415, p.112729_1 - 112729_16, 2023/12

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

JAEA Reports

Segregation of components of molten core oxidic materials using cold crucible induction heating technique (Joint research)

Sudo, Ayako; M$'e$sz$'a$ros, B.*; Sato, Takumi; Nagae, Yuji

JAEA-Research 2023-007, 31 Pages, 2023/11

JAEA-Research-2023-007.pdf:3.61MB

For the criticality assessment of fuel debris generated by the accident in Fukushima Daiichi Nuclear Power Station, understanding of the elemental localization in fuel debris is important. Especially, the distribution of Fe and Gd, which may behave as potential neutron absorber materials in the fuel debris, is of particular important from the viewpoint of nuclear criticality safety. To investigate the localization tendency of Gd and Fe in molten core materials during solidification progress, liquefaction/solidification tests on core materials containing UO$$_{2}$$, ZrO$$_{2}$$, FeO, Gd$$_{2}$$O$$_{3}$$, and simulated fission products (MoO$$_{3}$$, Nd$$_{2}$$O$$_{3}$$, SrO, and RuO$$_{2}$$) and concrete (SiO$$_{2}$$, Al$$_{2}$$O$$_{3}$$, and CaO) were performed using cold crucible induction heating technique. During the test, the molten core materials gradually subsided and solidified from the bottom to the top of the melt. Elemental analysis showed that Fe content in the inner region increased approximately up to 3.4 times that in the bottom region. The concentration of Fe into the inner region was observed in all the samples regardless of the initial FeO composition, cooling rates, and phase separation. This suggests that FeO may be concentrated into the low temperature region, where the melt solidified later. In contrast, Gd content in the bottom region increased approximately up to 2.6 times that in the inner region. The concentration of Gd into the bottom region was observed when the initial Gd$$_{2}$$O$$_{3}$$ content was higher than 1 at.%. This suggests that Gd$$_{2}$$O$$_{3}$$ may be concentrated into the earlier solidified region. On the other hand, no significant localization was observed on the simulated fission products.

Journal Articles

Axial variations of oxide layer growth and hydrogen uptake of BWR fuel claddings under steam starvation conditions

Sakamoto, Kan*; Adachi, Mika*; Tokushima, Kazuyuki*; Aomi, Masaki*; Shibata, Hiroki; Nagae, Yuji; Kurata, Masaki

Zirconium in the Nuclear Industry; 20th International Symposium (ASTM STP 1645), p.411 - 432, 2023/11

Journal Articles

Summary results of subsidy program for the "Project of Decommissioning, Contaminated Water and Treated Water Management (Development of Analysis and Estimation Technologies for Characterization of Fuel Debris (Development of Estimation Technologies of RPV Damaged Condition, etc.) in 2022JFY

Yamashita, Takuya; Shimomura, Kenta; Nagae, Yuji; Yamaji, Akifumi*; Mizokami, Shinya; Mitsugi, Takeshi; Koyama, Shinichi

Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 53 Pages, 2023/10

JAEA performed the subsidy program for the "Project of Decommissioning, Contaminated Water and Treated Water Management (Development of Analysis and Estimation Technologies for Characterization of Fuel Debris (Development of Estimation Technologies of RPV Damaged Condition, etc.) in 2022JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning, Contaminated Water and Treated Water Management.

Journal Articles

Numerical simulation method using a Cartesian grid for oxidation of core materials under steam-starved conditions

Yamashita, Susumu; Sato, Takumi; Nagae, Yuji; Kurata, Masaki; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 60(9), p.1029 - 1045, 2023/09

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Study on chemical interaction between UO$$_{2}$$ and Zr at precisely controlled high temperatures

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

Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06

 Times Cited Count:2 Percentile:65.72(Nuclear Science & Technology)

Interaction tests between UO$$_{2}$$ and Zr were performed at precisely controlled high temperatures between 1840 and 2000 $$^{circ}$$C to understand the interaction mechanism in detail. A Zr rod was inserted in a UO$$_{2}$$ crucible and then heat-treated at a fixed temperature in Ar-gas flow for 10 min. After heating in the range of 1890 to 1930 $$^{circ}$$C, the Zr rod was deformed to a round shape, in which the post-analysis detected the significant diffusion of U into the Zr region and the formation of a dominant $$alpha$$-Zr(O) matrix and a small amount of U-Zr-O precipitates. The abrupt progress of liquefaction was observed in the sample heated at around 1940 $$^{circ}$$C or higher. The higher oxygen concentration in the $$alpha$$-Zr(O) matrix suppressed the liquefaction progress, due to the variation in the equilibrium state. The U-Zr-O melt formation progressed by the selective dissolution of Zr from the matrix, and the selective diffusion of U could occur via the U-Zr-O melt.

JAEA Reports

Collection of strength characteristic data used for analysis evaluation in reactor pressure vessel and in-core structures in accident

Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji

JAEA-Data/Code 2022-012, 270 Pages, 2023/03

JAEA-Data-Code-2022-012.pdf:38.25MB

In a light water reactor, which is a commercial nuclear power plant, a severe accident such as loss of cooling function in the reactor pressure vessel (RPV) and exposure of fuel rods due to a drop in the water level in the reactor can occur when a trouble like loss of all AC power occurs. In the event of such a severe accident, the RPV may be damaged due to in-vessel conditions (temperature, molten materials, etc.) and leakage of radioactive materials from the reactor may occur. Verification and estimation of the process of RPV damage, molten fuel debris spillage and expansion, etc. during accident progression will provide important information for decommissioning work. Possible causes of RPV failure include failure due to loads and restraints applied to the RPV substructure (mechanical failure), failure due to the current eutectic state of low-melting metals and high-melting oxides with the RPV bottom members (failure due to inter-material reactions), and failure near the melting point of the structural members at the RPV bottom. Among the failure factors, mechanical failure is verified by numerical analysis (thermal hydraulics and structural analysis). When conducting such a numerical analysis, the heat transfer properties (thermal conductivity, specific heat, density) and material properties (thermal conductivity, Young's modulus, Poisson's ratio, tensile, creep) of the materials (zirconium, boron carbide, stainless steel, nickel-based alloy, low alloy steel, etc.) constituting the RPV and in-core structures to near the melting point are required to evaluate the creep failure of the RPV. In this document, we compiled data on the properties of base materials up to the melting point of each material constituting the RPV and in-core structures, based on published literature. In addition, because welds exist in the RPV and in-core structures, the data on welds are also included in this report, although they are limited.

Journal Articles

Thermodynamic analysis for solidification path of simulated ex-vessel corium

Sato, Takumi; Nagae, Yuji; Kurata, Masaki; Quaini, A.*; Gu$'e$neau, C.*

CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 79, p.102481_1 - 102481_11, 2022/12

 Times Cited Count:0 Percentile:0.00(Thermodynamics)

Journal Articles

X ray spectroscopy on $$Xi$$$$^-$$ atoms (J-PARC E03, E07 and future)

Yamamoto, Takeshi; Fujita, Manami; Gogami, Toshiyuki*; Harada, Takeshi*; Hayakawa, Shuhei*; Hosomi, Kenji; Ichikawa, Yudai; Ishikawa, Yuji*; Kamata, K.*; Kanauchi, H.*; et al.

EPJ Web of Conferences, 271, p.03001_1 - 03001_5, 2022/11

Journal Articles

An Investigation of the microstructure and phase composition of the Zr bearing metallic debris in a bypass channel of a BWR fuel after the exothermic reaction in the CLADS-MADE-04 test

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

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10

The CLADS-MADE-04 is the next test in the series aiming at understanding of the melt propagation behaviour in the lower core region. In this contribution, recent results of the post-test analysis including microstructure of metallic debris investigated by Electron Probe Micro Analyzer (EPMA) are discussed. During the test, melting of the control blade happened with sudden wave of strong heat release relatively slowly (several cm/min) spread from the hottest area downwards along the degrading control blade and channel box consuming the walls made of Zircaloy-4. A significant damage happened with the sample supporting plate as well. The investigation of microstructure of such metallic debris would allow understanding of a mechanism of enhanced local core degradation. The nature of strong heat release and possibility of spreading to the surrounding materials is to be confirmed after thorough phase identification by EPMA. The difference between Fe-B eutectic debris and Zr-Fe eutectic debris will be outlined. It is especially important for understanding of the lower core plate melt-through and a possibility of a Zr-Fe molten material progression into the lower plenum.

Journal Articles

Thermodynamic evaluation on solidification path for U-Zr-Fe-O corium

Ohgi, Hiroshi; Nagae, Yuji; Kurata, Masaki

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10

Journal Articles

BWR lower head penetration failure test focusing on eutectic melting

Yamashita, Takuya; Sato, Takumi; Madokoro, Hiroshi; Nagae, Yuji

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

 Times Cited Count:1 Percentile:19.69(Nuclear Science & Technology)

Journal Articles

Ten years of Fukushima Dai-ichi post-accident research on the degradation phenomenology of the BWR core components

Pshenichnikov, A.; Shibata, Hiroki; Yamashita, Takuya; Nagae, Yuji; Kurata, Masaki

Journal of Nuclear Science and Technology, 59(3), p.267 - 291, 2022/03

 Times Cited Count:3 Percentile:24.93(Nuclear Science & Technology)

Journal Articles

Raman investigation of the CLADS-MADE-02 test debris to confirm the mechanism of the volatile and non-volatile boron compounds formation

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

Proceedings of TopFuel 2021 (Internet), 12 Pages, 2021/10

Journal Articles

A BWR control blade degradation observed in situ during a CLADS-MADE-02 test under Fukushima Dai-Ichi Unit 3 postulated conditions

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

Journal of Nuclear Science and Technology, 58(9), p.1025 - 1037, 2021/09

 Times Cited Count:4 Percentile:47.47(Nuclear Science & Technology)

Journal Articles

Features of a BWR neutron absorber melt relocation in an oxidative environment during the CLADS-MADE-02 test

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

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 7 Pages, 2021/08

229 (Records 1-20 displayed on this page)