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Journal Articles

Evaluation of the effect of spent fuel layout on SFP cooling with MAAP5.04

Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Kaji, Yoshiyuki; Nemoto, Yoshiyuki

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

After the accident of Fukushima-unit 1 Nuclear Power Plant, Japanese utilities are newly requested by regulatory body to take prompt measures to enhance the safety of spent fuel pool. The most important objective of this new Japanese standards of regulation is keeping a water level in a Spent Fuel Pool (SFP) under any situations in order to prevent fuel failures due to increase of fuel temperature and to avoid the occurrence of re-criticality accidents. The utilities are considered to install several kinds of safety measures for SFP. For example, a spray injection and an alternate water injection to keep pool water level, and a spent fuel layout, such as 1 by 4, 1 by 8, checkerboard to enhance cooling of the spent fuel in SFP. The objective of the present study is to investigate the effect of spent fuel layout on SFP cooling with MAAP5.04.

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

Analysis of transport behaviors of cesium and iodine in VERDON-2 experiment for chemical model validation

Shiotsu, Hiroyuki; Ito, Hiroto*; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Journal Articles

Computational fluid dynamics analysis for hydrogen deflagration tests at ENACCEF2 facility

Trianti, N.; Sato, Masatoshi*; Sugiyama, Tomoyuki; Maruyama, Yu

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

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

Free convective heat transfer experiment to validate air-cooling performance analysis of fuel debris

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

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

Development of numerical simulation method for small particles behavior in two-phase flow by combining interface and Lagrangian particle tracking methods

Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Journal Articles

Numerical simulation of thermal hydraulics around a beam window in accelerator-driven system

Yamashita, Susumu; Yoshida, Hiroyuki

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

To investigate detailed flow behaviors around the beam window of accelerator driven system (ADS), large scale simulation for unsteady thermal hydraulics around the beam window was performed using JUPITER. The input data, such as the beam window and nozzle, is designed as accurate as possible using the computer aided design software. As a result, the flow pattern around the beam window is quite different from previous results in which the steady flow is assumed. The flow pattern of the Lead-Bismuth Eutectic around the beam window and the exit of the nozzle were very complicated. According to complicated flow around those structures, the temperature distribution on the beam window is also complicated. Thus, it is confirmed that the complicated flow around structures will affect to the temperature distribution in the structures and the effect of flow field on the temperature must be evaluated.

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

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

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

Measurement of void fraction distribution in a 4$$times$$4 fuel bundle under high pressure condition for validation of two-phase CFD code

Nagatake, Taku; Shibata, Mitsuhiko; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki

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

In the Fukushima Daiichi Nuclear Power Plant accident, reactor cores were cooled by natural circulation due to pump trip. To investigate the accident progress of the Fukushima Daiichi Nuclear Power Plant, it is important to understand the thermal hydraulic behavior in reactor cores including fuel bundles. Flow rate inside cores was relatively low in the natural circulation conditions, then, thermal-hydraulic behavior in the fuel bundles was different from that in the normal operating conditions. To evaluate thermal hydraulic behavior under the accidental conditions, we are developing the numerical simulation codes named TPFIT and ACE3D. These codes are based on two-phase computational fluid dynamics and can simulate the two-phase flow inside fuel bundles including low flow rate condition. Before applying these codes to the thermal-hydraulic behavior, the applicability of these codes must be confirmed. Then, in this study, in order to obtain a validation data for TPFIT and ACE3D code, thermal hydraulic experiment was performed by using test section with a simulated fuel bundle with 4$$times$$4 unheated rods. In this simulated fuel bundle, there were wire mesh sensors, and void fraction distribution data inside the simulated fuel bundle under high pressure condition (max. 2.6 MPa) was obtained. The one of the advantage of wire mesh sensor is that a void fraction distribution of cross section at the same time can be measured. In this paper, void fraction distribution of two-phase flow in a simulated fuel bundle under high pressure condition are reported.

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.

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