Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 11232

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Characterization of high-temperature nuclear fuel-coolant interactions through X-ray visualization and image processing

Johnson, M.*; Journeau, C.*; Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji

Annals of Nuclear Energy, 151, p.107881_1 - 107881_13, 2021/02

High-resolution X-ray imaging was employed at the JAEA MELT facility to visualize a kilogram-scale interaction between a jet of high temperature molten stainless steel and sodium. A novel software, SPECTRA, has been developed for the quantitative characterization of jet quenching and fragmentation. Tracking and 3D reconstruction of the melt phase traversing the imaging window enabled the detection of 72% of the debris mass recovered post-experiment. The rebounding of melt fragments confirmed a solid outer crust at the melt-coolant interface, while a thermal fragmentation event induced rapid vapor expansion. Jet fragmentation is best explained by the vaporization of coolant entrained within the melt jet generating an internal over-pressure sufficient for fragmentation of the crust. Thermal fragmentation produced a bimodal debris size distribution of coarse jet shells and finer fragments.

Journal Articles

Numerical simulation of heat transfer behavior in EAGLE ID1 in-pile test using finite volume particle method

Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji

Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01

Journal Articles

Experimental study on secondary droplets produced during liquid jet impingement onto a horizon solid surface

Zhan, Y.*; Kuwata, Yusuke*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi

Experimental Thermal and Fluid Science, 120, p.110249_1 - 110249_12, 2021/01

Journal Articles

Oxygen self-diffusion in near stoichiometric (U,Pu)O$$_{2}$$ at high temperatures of 1673-1873 K

Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*

Journal of Nuclear Materials, 542, p.152472_1 - 152472_7, 2020/12

 Times Cited Count:0

The oxygen self-diffusion coefficients in near stoichiometric (U,Pu)O$$_{2}$$ at high temperatures were successfully measured by thermogravimetry combined with the oxygen isotope exchange method. The activation energy for oxygen diffusion in the stoichiometric composition of (U,Pu)O$$_{2}$$ was evaluated from experimental data, and the value was determined to be 248 kJ/mol. In addition, the defect migration energies of (U,Pu)O$$_{2 pm x}$$ were derived, and the oxygen self-diffusion coefficients were evaluated using these. As a result, good agreement was found between the experimental data and the oxygen self-diffusion coefficients calculated using the defect migration energies.

Journal Articles

Experiments of self-wastage phenomena elucidation in steam generator tube of sodium-cooled fast reactor

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12

Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.

Journal Articles

Evaluation of breach characteristics of fast reactor fuel pins during steady state irradiation

Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi

Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12

The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.

Journal Articles

Safety enhancement approach against external hazard on JSFR reactor building

Yamamoto, Tomohiko; Kato, Atsushi; Chikazawa, Yoshitaka; Hara, Hiroyuki*

Nuclear Technology, 206(12), p.1875 - 1890, 2020/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

This paper gives a detailed evaluation of the countermeasures for the external hazards and severe accidents that could impact the 2010 JSFR design building by lessons learned from the Fukushima Daiichi nuclear power plant (Fukushima I NPP) accident.

Journal Articles

Next generation reactor development and current status

Kamide, Hideki

Karyoku Genshiryoku Hatsuden, 71(11), p.638 - 648, 2020/11

Development status of next generation reactors is outlined mainly for sodium cooled fast reactor developed in Japan Atomic Energy Agency (JAEA). Development strategy in Japan, status of the advanced reactor developments including SMR over the world, and also results of research and development in JAEA are explained.

Journal Articles

Internal event level-1 PRA for sodium-cooled fast reactor considering safety measures of defense-in-depth level 1 to 3

Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya; Yamano, Hidemasa

Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 12 Pages, 2020/11

The objective of this study is to evaluate the occurrence frequency of accident sequences which may lead to core damage if provisions in defense in depth (DiD) level 1 to 3 are the only safety measures. For this objective, the existing safety measures in this SFR are categorized into those for the DiD level 1-3 and those for the DiD level 4. The safety measures for the DiD level 1-3 are as follows; (1) main reactor shutdown system, (2) double boundary structure in the primary main and auxiliary cooling system and the reactor vessel, which maintain the reactor coolant level sufficient for coolant circulation in the primary main cooling system, (3) decay heat removal in a forced circulation mode. Accident sequences are categorized into typical SFR-specific groups and station blackout (SBO) in this study. The SFR-specific groups are unprotected loss of flow, unprotected transient over power, unprotected loss of heat sink, loss of reactor level, and protected loss of heat sink (PLOHS). The occurrence frequency of these accident sequence groups was quantified to identify major contributors. As the result, PLOHS excluding SBO was indicated as the dominant contribution of 80% or more in the all accident sequence groups and the annual occurrence frequency of the PLOHS was 1.0E-4 order of magnitude. For the PLOHS, loss of offsite power (LOOP) was indicated as major contribution of 30% in initiating events. In the accident sequences of the PLOHS initiated from LOOP, a dominant sequence was combination of common cause failure of primary pumps in the main cooling system and failure-to-start of the auxiliary cooling system after LOOP. The second dominant contribution (15% or more) in the all accident sequence groups is PLOHS in SBO (i.e., decay heat removal failure due to SBO). Each of the other accident sequence groups was 1%.

Journal Articles

Numerical simulation of the solid particle sedimentation and bed formation behaviors using a hybrid method

Sheikh, M. A. R.*; Liu, X.*; Matsumoto, Tatsuya*; Morita, Koji*; Guo, L.*; Suzuki, Toru*; Kamiyama, Kenji

Energies (Internet), 13(19), p.5018_1 - 5018_15, 2020/10

 Times Cited Count:0

Journal Articles

Measurement time and result of apatite FT density using FT automatic measuring device

Shimada, Koji; Sueoka, Shigeru

Fisshion, Torakku Nyusureta, (33), p.19 - 21, 2020/10

no abstracts in English

Journal Articles

Phase-field model for crystallization in alkali disilicate glasses; Li$$_2$$O-2SiO$$_2$$, Na$$_2$$O-2SiO$$_2$$ and K$$_2$$O-2SiO$$_2$$

Kawaguchi, Munemichi; Uno, Masayoshi*

Journal of the Ceramic Society of Japan, 128(10), p.832 - 838, 2020/10

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

This study developed phase-field method (PFM) technique in oxide melt system by using a new mobility coefficient ($$L$$). The crystal growth rates ($$v_0$$) obtained by the PFM calculation with the constant $$L$$ were comparable to the thermodynamic driving force in normal growth model. The temperature dependence of the $$L$$ was determined from the experimental crystal growth rates and the $$v_0$$. Using the determined $$L$$, the crystal growth rates ($$v$$) in alkali disilicate glasses, Li$$_2$$O-2SiO$$_2$$, Na$$_2$$O-2SiO$$_2$$ and K$$_2$$O-2SiO$$_2$$ were simulated. The temperature dependence of the $$v$$ was qualitatively and quantitatively so similar that the PFM calculation results demonstrated the validity of the $$L$$. Especially, the $$v$$ obtained by the PFM calculation appeared the rapid increase just below the thermodynamic melting point ($$T_{rm m}$$) and the steep peak at around $$T_{rm m}$$-100 K. Additionally, as the temperature decreased, the $$v$$ apparently approached zero ms$$^-1$$, which is limited by the $$L$$ representing the interface jump process. Furthermore, we implemented the PFM calculation for the variation of the parameter $$B$$ in the $$L$$. As the $$B$$ increased from zero to two, the peak of the $$v$$ became steeper and the peak temperature of the $$v$$ shifted to the high temperature side. The parameters $$A$$ and $$B$$ in the $$L$$ increased exponentially and decreased linearly as the atomic number of the alkali metal increased due to the ionic potential, respectively. This calculation revealed that the $$A$$ and $$B$$ in the $$L$$ were close and reasonable for each other.

Journal Articles

An Investigation on the control rod homogenization method for next-generation fast reactor cores

Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.92 - 96, 2020/10

Journal Articles

Validation study of finite element thermal-hydraulics analysis code SPIRAL to a large-scale wire-wrapped fuel assembly at low flow rate condition

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10

A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.

Journal Articles

Evaluation of the characteristics of metal nitrate aqueous solutions by microwave heating and the morphologies of synthesized metal oxide powders

Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Suzuki, Masahiro; Fukasawa, Tomonori*; Fukui, Kunihiro*

Funtai Kogakkai-Shi, 57(9), p.485 - 494, 2020/09

In the spent fuel reprocessing process, a mixed solution of uranyl nitrate and plutonium nitrate is converted into mixed oxide powder by the microwave heating. To evaluate the applicability to the industrial-scale and acquire the characteristics data of the microwave heating denitration of various metal nitrate aqueous solutions based on the knowledge studied in the development of laboratory-scale basic experiments, the microwave heating characteristics and metal oxide powder properties were investigated using cerium nitrate, cobalt nitrate and copper nitrate aqueous solutions. The progress rate of the denitration reaction was depended on the position, and the denitration reaction proceeded faster at the periphery than at the center. The morphologies of the synthesized products were porous and hard dry solid with cerium nitrate aqueous solution, foamed dry solid with cobalt nitrate aqueous solution, and powdery particles with copper nitrate aqueous solution. The denitration ratio and average particle size of the synthesized products increased in the order of the cerium nitrate aqueous solution, the cobalt nitrate aqueous solution, and the copper nitrate aqueous solution. The numerical simulations revealed that the periphery of the bottom surface of the metal nitrate aqueous solution was heated by microwaves. This results consistent with the experimental results in which the denitration reaction started from the periphery of the metal nitrate aqueous solution.

Journal Articles

The Evaluation of the properties of the collision-plate-type jet mill for dry recycling of MOX powder

Kawaguchi, Koichi; Segawa, Tomoomi; Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Ishii, Katsunori

Funtai Kogakkai-Shi, 57(9), p.478 - 484, 2020/09

A collision plate type jet mill is assumed to be a pulverizer that can control the particle size for nuclear fuel fabrication. The collision plate type jet mill consists of two modules, a classifier and a mill chamber. Coarse component of powder is cycled in the equipment and finally pulverized into objective particle size. In this report, simulated crushed powders were classified and pulverized step by step, and particle size distribution were compared. The collision plate type jet mil can produce objective size particles with low overgrinding.

Journal Articles

Innovation for flexible use of nuclear power in JAEA

Kamide, Hideki; Shibata, Taiju

NREL/TP-6A50-77088 (Internet), p.35 - 38, 2020/09

Journal Articles

Kinetic study on eutectic melting process between boron carbide and stainless steel in sodium-cooled fast reactor

Kikuchi, Shin; Sakamoto, Kan*; Takai, Toshihide; Yamano, Hidemasa

Nippon Kikai Gakkai 2020-Nendo Nenji Taikai Koen Rombunshu (Internet), 4 Pages, 2020/09

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) as control rod element and stainless steel (SS) as control rod cladding or related structure may occur. Thus, behavior of B$$_{4}$$C-SS eutectic melting is one of the phenomena to evaluate the core disruptive accidents in SFR. In order to clarify the kinetic feature of B$$_{4}$$C-SS eutectic melting process in the interface, the thinning test for SS crucibles using the pellets of B$$_{4}$$C or SS with low B$$_{4}$$C concentration were performed to obtain the rate constant with dependence of B$$_{4}$$C concentration against SS. It was found that the rate constants of eutectic melting between SS and SS low B$$_{4}$$C concentration were smaller than that of B$$_{4}$$C-SS in the high temperature range. Besides, the rate constant of eutectic melting between SS and B$$_{4}$$C containing SS became small when decreasing the B$$_{4}$$C concentration against SS.

Journal Articles

Assessment of nuclear simulation credibility

Tanaka, Masaaki; Nakada, Kotaro*; Kudo, Yoshiro*

Nippon Kikai Gakkai-Shi, 123(1222), p.26 - 29, 2020/09

In the nuclear engineering, simulations are used in radiation, thermal hydraulic, chemical reaction, and structural fields, and the integrated fields thereof, to be applied to the design, construction and operation of nuclear facilities. This article describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements described in the "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)". And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.

Journal Articles

Oxide dispersion strengthened steels

Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi

Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08

Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.

11232 (Records 1-20 displayed on this page)