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Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10
Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30%(9MW). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.
Moriyama, Kiyofumi; Nakamura, Hideo; Maruyama, Yu
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 18 Pages, 2004/10
A steam explosion simulation code JASMINE is under development at JAERI for the assessment of steam explosion impacts on the integrity of containment vessel during severe accidents in light water reactors. Selected alumina and corium steam explosion experiments, KROTOS-44, 42, 37 and FARO-L33 were simulated with JASMINE code. The experimentally observed difference of the steam explosion intensity with the two materials, alumina and corium, was reproduced in the simulations without changing the model parameters related to the fine fragmentation process, but based on the difference in the premixing behavior predicted by the simulations. The simulation of corium experiments showed more fraction of the melt droplets frozen during premixing, as well as more void fraction, and those two points were likely to be the primary reasons of weak interactions in corium experiments.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Yonomoto, Taisuke; Akie, Hiroshi; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao; Iwamura, Takamichi
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10
Reduced-Moderation Water Reactor (RMWR) is a light-water cooled high-conversion reactor that is being developed by JAERI with collaboration from the Japanese industries. Since RMWR utilizes the highly enriched plutonium, the safety concern for RMWR includes the possibility of recriticality during severe accidents as is the case with the liquid metal cooled fast breeder reactor. In order to clarify this concern, characteristics of severe accidents of RMWR are analyzed in this study. The results obtained so far indicate that (1) the mechanical impact of recriticality in the core, if occurs, is supposed to be insignificant due to the absence of water, (2) the mixture of the fuel and cladding debris in the lower plenum does not cause recriticality when they are well mixed and distributed flatly, and (3) if requires, the installation of neutron-absorption material with realistic geometry can effectively prevent recriticality in the lower plenum even for the conservatively-assumed spherical accumulation of core debris.
Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 17 Pages, 2004/10
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Kamiyama, Kenji
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 0 Pages, 2004/10
In this study, model improvement of SIMMER-III code is presented which is based on the knowledge of metallurgy area consisting of the microstructure observation of solodofoed materials. The formation of supercooling layer in the vicinity of wall surface and the imperfect contact of melt to the wall were confirmed. Some applications to freezing tests showed that the temperature of supercooling layer dominated the rate of bulk enthalpy loss and it was specific to each material. Thus, in order to increase the applicability and accuracy of SIMMER-III code for freezing phenomena in CDA, a new correlation was proposed in this study to predict the temperature of supercooling layer in the vicinity of wall.
Okano, Yasushi
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 112 Pages, 2004/10
Commercial computational fluid dynamic program is taken up to be employed for nuclear thermal-hydraulic applications due to the advantages in high speed solution and easy-to-use operation. The principal objective of this manuscript is evaluating the numerical simulation accuracy of the Fluent, on single phase multi-dimensional thermal hydraulic problems.
Kamide, Hideki; Kimura, Nobuyuki; Igarashi, Minoru; Hayashi, Kenji
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 0 Pages, 2004/10
Experimental Study was carried out for evaluation of thermal hydraulics of thermal striping. Parallel-triple jet along a wall and mixing tee experiments and experimenatl analyses using DINUS-3 code were performed. Decay of fluctuation intensity near the wall was found. Calculated results of temperature and velocity fields were in good agreement with experimental data.
Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00
Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.
Takata, Takashi; Yamaguchi, Akira
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), P. 146, 2004/00
Sensitivity study of Sodium-Water Reaction (SWR), which will occur in a steam generator of a liquid sodium cooled fast reactor when a heat transfer tube is failed, have been carried out using the multi-dimensional and multi-phase analysis code SERAPHIM. An influence of the interfacial area density and the sodium hydroxide (NaOH) evaporation on the SWR phenomena is investigated. As a result, it is concluded that the interfacial area density affects the high temperature region strongly. It is also investigated that the evaporation of NaOH influences on the maximum temperature in the SWR.