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Journal Articles

The Effects of plutonium content and self-irradiation on thermal conductivity of mixed oxide fuel

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Kato, Masato; Saito, Kosuke; Uno, Masayoshi*

Nuclear Technology, 205(3), p.474 - 485, 2019/03

This study evaluated the effects of plutonium content and self-irradiation on the thermal conductivity of mixed-oxide (MOX) fuel. Samples of UO$$_{2}$$ fuel and various MOX fuels were tested. The MOX fuels had a range of plutonium contents, and some samples were stored for 20 years. The thermal conductivity of these samples was determined from thermal diffusivity measurements taken via laser flash analysis. Although the thermal conductivity decreased with increasing plutonium content, this effect was slight. The effect of self-irradiation was investigated using the stored samples. The reduction in thermal conductivity caused by self-irradiation depended on the plutonium content, its isotopic composition, and storage time. The reduction in thermal conductivity over 20 years' storage can be predicted from the change of lattice parameter. In addition, the decrease in thermal conductivity caused by self-irradiation was recovered with heat treatment, and recovered almost completely at temperatures over 1200 K. From these evaluation results, we formulated an equation for thermal conductivity that is based on the classical phonon-transport model. This equation can predict the thermal conductivity of MOX fuel thermal conductivity by accounting for the influences of plutonium content and self-irradiation.

Journal Articles

Austenite-based stainless steel irradiation behavior of the precipitate and void swelling

Inoue, Toshihiko; Sekio, Yoshihiro; Watanabe, Hideo*

Materia, 58(2), P. 92, 2019/02

For the evaluation of irradiated segregation behavior, Austenite-based stainless steel for the fast reactor, during irradiation was evaluated by utilizing TIARA facility (Irradiate temperature: 600 $$^{circ}$$C, Dose: 100 dpa) was observed by analytical electron microscope (JEM-ARM20FC). As a result of observation, the large-size void is observed in irradiation area, and MX segregation (containing Niobium) is not observed. In un-irradiation area the MX segregation is observed. And it is observed conspicuously that Nickel is segregation on the void surface. By the latest high-performance TEM utilization, these phenomenon are able to visualize. It is expected for the clarification of the irradiation damage and mechanism of void swelling, by the analyzing these phenomenon utilization with the latest high-performance TEM utilization.

Journal Articles

Development of creep property equations of 316FR stainless steel and Mod.9Cr-1Mo steel for sodium-cooled fast reactor to achieve 60-year design life

Onizawa, Takashi; Hashidate, Ryuta

Mechanical Engineering Journal (Internet), 6(1), p.18-00477_1 - 18-00477_15, 2019/02

Aiming at enhancing its economic competitiveness and reducing radioactive waste, JAEA has proposed an attractive plant concept and made great efforts to demonstrate the applicability of some innovative technologies to the plant. One of the most practical means is to extend the design life to 60 years. Accordingly, the material strength standards set by JSME have to be extended from 300,000 to 500,000 hours but this extension requires more precise estimation of creep rupture strength and creep strain of the materials in the long term. This paper describes the development of creep property equations of 316FR stainless steel and Mod.9Cr-1Mo steel considering changes in creep mechanisms at high temperatures in the long term based on evaluations of long-term creep properties of the materials. The creep property equations developed in this study will provide more precise estimation of the creep properties in the long term than the present creep property equations of JSME.

Journal Articles

Challenge next-generation nuclear system; Development of oxide dispersion strengthened ferritic steel

Otsuka, Satoshi; Kaito, Takeji

Enerugi Rebyu, 39(1), p.44 - 46, 2019/01

For performance improvement of next-generation nuclear system such as fast reactor, it has been expected to develop advanced material resistant to severe in-reactor environment (i.e. high-dose neutron irradiation at high-temperature). Japan Atomic Energy Agency (JAEA) has been developing Oxide Dispersion Strengthened (ODS) ferritic steel for long life fuel cladding tube of fast reactor. Application of ODS ferritic steel to fast reactor fuel can extend the fuel life time twice or more as long as the fuel with conventional cladding tube (i.e. modified SUS316), thus reducing fuel exchange frequency and fuel cost. It can be adaptable to high-temperature plant operation, which is favorable for improvement of power generation efficiency. This paper interprets the development of ODS ferritic steel cladding tube for sodium-cooled fast reactor, which has been led by JAEA for dozens of years.

Journal Articles

Development and validation of SAS4A code and its application to analyses on severe flow blockage accidents in a sodium-cooled fast reactor

Fukano, Yoshitaka

Journal of Nuclear Engineering and Radiation Science, 5(1), p.011001_1 - 011001_13, 2019/01

Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code clarified that the conclusion in the past study was almost same as that in this study. Furthermore SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.

Journal Articles

Influences of pore and particle sizes of CMPO/SiO$$_{2}$$-P adsorbent on extraction chromatography process

Watanabe, So; Sano, Yuichi; Sanda, Shuhei*; Sakurai, Shota*; Arai, Tsuyoshi*

Nippon Ion Kokan Gakkai-Shi, 30(1), p.8 - 16, 2019/01

Journal Articles

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki

Nuclear Technology, 205(1-2), p.119 - 127, 2019/01

To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.

Journal Articles

Performance evaluation of eddy current flowmeter in Monju

Aizawa, Kosuke; Chikazawa, Yoshitaka; Morohashi, Yuko

Journal of Nuclear Science and Technology, 55(12), p.1393 - 1401, 2018/12

 Percentile:100(Nuclear Science & Technology)

Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-initiated accident in the early stage and to understand the reactor's thermal hydrodynamic performance. Japan Atomic Energy Agency has developed the eddy current flowmeter in practical use and installed 34 of them in the upper core structure of fast breeder reactor, Monju. This report presents data obtained by using the flowmeters in Monju. We observed high linearity between each of the flowmeter's signal intensity and the primary sodium's flow rate under 10-100% flow rate condition. The fluctuation of flow rate observed by the flowmeters was below 0.2 m/s which is 5% of the time-averaged velocity under a rated condition. These experimental results show that the eddy current flowmeter is an effective tool to detect the changes in relative flow rate.

Journal Articles

Development of safety design guideline of structures, systems and components of Generation-IV Sodium-cooled Fast Reactor

Okano, Yasushi

Nippon Genshiryoku Gakkai-Shi, 60(12), p.764 - 769, 2018/12

JAEA has developed the second safety design guidelines report, "Safety Design Guidelines on Structures, Systems and Components" for Generation-IV SFR system, following the previously published SFR Safety Design Criteria and the first SFR Safety Design Guidelines report and with the reviews by a technical committee under Atomic Energy Society of Japan. This article explains about 14 key points on reactor core system, coolant system, and containment system and also current consistency of international SFR designs to the safety design guidelines.

Journal Articles

Droplet generation during liquid jet impingement onto a horizontal plate

Zhan, Y.*; Oya, Naoki*; Enoki, Koji*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi

Experimental Thermal and Fluid Science, 98, p.86 - 94, 2018/11

 Percentile:100(Thermodynamics)

Journal Articles

Multi-dimensional numerical investigation of sodium spray combustion; Benchmark analysis of SNL T3 experiment

Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Denman, M. R.*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11

no abstracts in English

Journal Articles

Numerical analysis of core disruptive accident in a metal-fueled sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

Based on the event tree analysis, the present numerical analyses investigated the capability of fuel discharge through the one-dimensional single fuel assembly geometry and the two-dimensional geometry of a CRGT channel with neighboring fuel assemblies. The single fuel assembly analyses showed that the fuel blockage formed in the lower shielding region because fuel solidified by contacting with cold sodium in case of no fission gas release. On the assumption that fission gas was released, the molten fuel successfully relocated below the core. The next analyses using the CRGT channel indicated a significant fuel discharge through the CRGT channel. This is because the fuel temperature was still high just after the CRGT wall failure and sodium in the CRGT channel was quickly voided just after the ingress of a small amount of molten fuel.

Journal Articles

Mechanism of flashing phenomena by microwave heating and influence of high dielectric constant solution

Fujita, Shunya*; Abe, Yutaka*; Kaneko, Akiko*; Yuasa, Tomohisa*; Segawa, Tomoomi; Kato, Yoshiyuki; Kawaguchi, Koichi; Ishii, Katsunori

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

Mixed uranium oxide and plutonium oxide powder is produced from uranyl nitrate and plutonium nitrate mixed solution by the microwave heating denitration method in the spent fuel reprocessing process. Since the microwave heating method is accompanied by a boiling phenomenon, it is necessary to fully grasp the operating conditions in order to avoid flashing and spilling in the mass production of denitrification technology for the future. In this research, it was clarified that the heat transfer coefficient became lower as the dielectric constant increased. The dominant factor of the blowing up phenomena is supposed to be generation of the innumerable bubble rather than bubble's growth.

Journal Articles

Measurement of Velocity Field in Five Jets Water Test (FIWAT) for thermal striping in sodium-cooled fast reactor

Aizawa, Kosuke; Kobayashi, Jun; Tanaka, Masaaki; Kurihara, Akikazu; Ishida, Katsuji*; Nagasawa, Kazuyoshi*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 10 Pages, 2018/11

A conceptual design of an advanced loop type sodium cooled reactor has been carried out in the Japan Atomic Energy Agency (JAEA). Temperature fluctuation is caused by mixing of fluids at different temperature from the control rod channels and the core fuel assemblies, high cycle thermal fatigue may arise on the Core Instrument Plane (CIP) at bottom of the Upper Internal Structure (UIS). In JAEA, 1/3-scaled five jets water tests (FIWAT) have been performed in order to investigate thermal striping phenomena around the CIP. In this study, the velocity field was measured in the mixing area between the jet outlet and the bottom of the structure by using particle image velocimetry (PIV) to compare with the temperature fluctuation characteristics.

Journal Articles

Preliminary calculation on thermal stratification phenomena in the fundamental sodium experiment "SuperCAVNA"

Ezure, Toshiki; Nagasawa, Kazuyoshi*; Tanaka, Masaaki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11

To establish an evaluation method of thermal stratification in sodium-cooled fast reactors (SFRs), a benchmark exercise was performed for a sodium experiment (SuperCAVNA) with a rectangular test section and heated wall. This paper presents a preliminary result using three-dimensional finite differential code AQUA. The influences of mesh size for heat exchange and turbulence model are studied, and the calculation results were also compared to the experimental results in the literature. Then, the calculation results reproduced the thermal stratification in SuperCAVNA experiment. The position and the temperature gradient of the stratified surface also showed good agreement with the experimental result. The applicability of the numerical approach employed in this study for the evaluation of thermal stratification problem in SFRs was confirmed.

Journal Articles

Validation of three-dimensional finite-volume-particle method for simulation of liquid-liquid mixing flow behavior

Kato, Masatsugu*; Funakoshi, Kanji*; Liu, X.*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji

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

Journal Articles

Effect of porosity distribution on two-phase pressure drop in a packed bed

Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

In the evaluation of the in-place cooling which is for the residual core materials in the severe accident of sodium-cooled fast reactors, pressure loss of two-phase flow in debris bed is one of the important factors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, experiment to measure the pressure loss in a packed bed of non-homogeneous porosity distribution was conducted, and the Lipinski model was modified dividing the cross section to evaluate the pressure loss in it. As a result, it was confirmed that agreement of the experimental values with the values by modified Lipinski model was better than that with the original Lipinski model.

Journal Articles

Results of an out-of-pile experiment for fragmentation of a simulated molten core material discharged into a shallow sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11

In Core Disruptive Accidents of Sodium-cooled Fast Reactors, molten core material would be discharged through control rod guide tubes into the inlet coolant plenums beneath the rector cores. The inlet coolant plenums have quite limited heights and sodium inventories. Therefore, in the inlet plenums, molten core material with a jet-like shape would impinge on the bottom of the plenum before it breaks up into fragments. In this study, to clarify fragmentation behavior in a shallow sodium pool whose height and volume are so limited that jet impingement on the bottom is expected, an out-of-pile experiment discharging molten alumina into a sodium pool was conducted. Although a small amount of alumina agglomeration was found on the bottom plate (steel disk) installed in the sodium pool, most of the molten alumina was fragmented into debris particles. Results obtained in the present experiment suggest that molten core material is fragmented and quenched even in a shallow sodium pool.

Journal Articles

Advancement of numerical analysis method for tube failure propagation

Uchibori, Akihiro; Takata, Takashi; Yanagisawa, Hideki*; Li, J.*; Jang, S.*

Proceedings of 2018 ANS Winter Meeting and Nuclear Technology Expo; Embedded Topical International Topical Meeting on Advances in Thermal Hydraulics (ATH 2018) (USB Flash Drive), p.1289 - 1294, 2018/11

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.

Journal Articles

Characteristics of TPDN/SiO$$_{2}$$-P adsorbent for MA(III) recovery

Kofuji, Hirohide; Watanabe, So; Takeuchi, Masayuki; Suzuki, Hideya; Matsumura, Tatsuro; Shiwaku, Hideaki; Yaita, Tsuyoshi

Progress in Nuclear Science and Technology (Internet), 5, p.61 - 65, 2018/11

10794 (Records 1-20 displayed on this page)