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Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Yoshikawa, Akira; Seki, Yohji; Nishi, Hiroshi; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), p.645 - 649, 2010/05
This paper overviews the research and development activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing development of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m, 80 cycles with the coolant condition as DEMO, 15 MPa 300
C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been almost completed.
Ebara, Shinji*; Aoya, Yuta*; Sato, Tsukasa*; Hashizume, Hidetoshi*; Yuki, Kazuhisa*; Aizawa, Kosuke; Yamano, Hidemasa
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
In this study, pressure measurement test is conducted to find out the FIV characteristic features related to the elbow turbulent flow, using 1/7 scale experimental loop simulating the JSFR cold leg piping. As the first step of multielbow piping, pressure measurement for single elbow was performed. The same measurement procedure was taken as a 1/15 scale experiment to assess the influence originated from the different scale flow at the same Reynolds number.
Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
For steam generator with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important issues need researching. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube with a similar inner diameter as that in the designed SG. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15 MPa. Six models for the prediction of two-phase multiplier were evaluated. The results showed the Chisholm correlation and homogeneous model gave best predictions. Note that in the homogeneous model verification, the homogeneous model was only used in the friction loss calculation. In the calculation of void fraction, which is necessary for static head, drift flux model, instead of homogeneous model, was used.
Sakai, Takaaki; Kotake, Shoji; Aoto, Kazumi; Ito, Takaya*; Kamishima, Yoshio*; Oshima, Jun*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
JAEA is now conducting "Fast Reactor Cycle Technology Development (FaCT)" project for commercialization before 2050s. A demonstration reactor for Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since FY2007 to determine referential reactor specifications for the next stage of design work of licensing and construction study. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. In this paper, the current status of the conceptual design study for the demonstration reactor plant is summarized.
Furukawa, Tomohiro; Kato, Shoichi; Hirakawa, Yasushi; Kondo, Hiroo; Nakamura, Hiroo
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
The EVEDA lithium test loop is constructed at the Oarai Research and Development Center, Japan Atomic Energy Agency. Since lithium is specified as a dangerous substance by a Japanese law, the countermeasure which assumed the lithium combustion incident is indispensable. In this experimental study, the fire-extinguishing behavior of four kinds of fire extinguishers - dryness sand, pearlite, Natrex-L and Natrex-M - to burning lithium was examined. In addition, the effect of depth of lithium pool on the fire-extinguishing performance of the candidate fire extinguisher was investigated to determine the amount of the fire extinguisher placed at the EVEDA lithium test loop.
Naito, Morimasa; Kishi, Hirokazu; Fukuoka, Naomi; Yamada, Tsutomu*; Ishida, Hideaki*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 7 Pages, 2010/05
As an alternative grouting material for the geological repository of long-lived radioactive waste, the "Superfine Spherical silica Grout" (SFSG) material is developed using a fine spherical silica and a fine calcium hydroxide. The developed SFSG material takes an advantage of its smaller particle size distribution (max. 1 micron or less) than those of the cementitious materials, and also provides a low alkaline environment so as to reduce unfavorable effects on the long-term performance of geological disposal system. The SFSG is a mixture of the super fine silica powder, the superfine calcium hydroxide and additives such as superplasticizer. Some preliminary laboratory experiments were carried out to characterize its fundamental properties from the viewpoint of practical use for geological disposal, which is required to be equivalent with the conventional cementitious materials in terms of penetrability, strength, pH performance and workability.
Terada, Atsuhiko; Yan, X.; Hino, Ryutaro; Sato, Hiroyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 5 Pages, 2010/05
When a primary pipe of the HTGR ruptures, helium coolant gas in the reactor blows out into the reactor confinement structure and the reactor primary system depressurizes. Consequently, the core graphite structures may be oxidized by the air and the complicated natural convection of multi component gas mixtures with chemical reactions would take place inside the reactor. Hence, JAEA studies showed the air ingress phenomena in the depressurized reactor and proposed a new passive mechanism of sustained counter air diffusion (SCAD) that has been shown effective in preventing major air ingress through natural circulation in the reactor. In the present plan, JAEA will construct an experimental reactor mockup including reactor core, the SCAD system, pressure vessel, coaxial pipe and so on. The core is made of graphite or ceramics and heated by electric heaters to allow for test operation up to 1200C. Present status of these activities will be presented.
Takegami, Hiroaki; Terada, Atsuhiko; Onuki, Kaoru; Hino, Ryutaro
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
The Japan Atomic Energy Agency has been conducting R&D on thermochemical water-splitting Iodine-Sulfur (IS) process for hydrogen production. A concept of sulfuric acid decomposer was developed featuring a heat exchanger block made of SiC. Although knowing the strength of the SiC block is important for the reliability assessment, it is difficult to evaluate a large-scale ceramics structure without destructive test. In this study, a novel approach for strength estimation of SiC structure was proposed. Optimum value of the Weibull modulus was determined for evaluating the lowest strength estimation. The strength estimation line was developed by using the determined modulus. The validity of the line was verified by destructive test of SiC block model, which is small-scale model of the SiC block. The fracture strength of small-scale model satisfied the predicted strength of the model.
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: debris-bed coolability and metallic fuel freezing behavior. To analyze the debris-bed coolability, the ACRR-D10 in-pile experiments were selected. SIMMER-III well simulated the heat transfer mechanisms including conduction, boiling and channeling observed in the experiment. Metallic fuel may freeze onto the stainless steel (cladding or wrapper tube) together with eutectic formation during core disruption in a metallic-fueled reactor. The CAFE-UT2 experiment carried out using pure Uranium melt to investigate such phenomena was selected for the experimental analysis. In spite of no eutectic formation model in the SIMMER-III code, the calculated fuel penetration behavior was in good agreement with the experimental data.
Morita, Koji*; Zhang, S.*; Arima, Tatsumi*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; et al.
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized.
Takada, Shoji; Abe, Kenji; Inagaki, Yoshiyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
The high temperature isolation valve (HTIV) is a key component to assure the safety of a HTGR connected with a hydrogen production system. The structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed. Inner diameter of the top of seat was 445 mm. Numerical analysis was carried out to estimate temperature and thermal stress of metallic components by 3-dimensional finite element method. Numerical results showed that the temperature of the seat was simply decreased from the top to the root, and the thermal stress locally increased at the root. Thermal stress was lowered below the allowable limit 120 MPa by optimizing the structure. The thermal stress increased at the top of the seat. A creep analysis showed that a creep damage factor was limited below allowable limit during the start-up and shut-down during normal operation, as well as during the depressurization accident.
Matsuo, Yoichiro; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
Radioactive corrosion products are main cause of personal radiation exposure during maintenance with no breached fuel in FBR plants. CP is produced in the core region by activation of fuel cladding and sub-assembly wrappers, and they are transported to the primary circuit with sodium flow and deposited on the wall of the primary piping and components. In order to establish the techniques of radiation dose reduction for of personnel, program system for corrosion hazard evaluation code PSYCHE has been was developed. The PSYCHE code is based on the solution-precipitation model. The density of each deposited CP and dose rate of primary coolant system in Monju was estimated by using the PSYCHE and QAD-CG code.
Miyahara, Shinya; Nishimura, Masahiro; Nakagiri, Toshio
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
Equilibrium partition coefficients were experimentally measured using "Transpiration method" for volatile fission products of cesium and iodine between liquid sodium pool and the inert cover gas. The objectives of the experiments are to : (1) Obtain the equilibrium partition coefficients of cesium and iodine at high temperature between 600 and 850 C and (2) Study the dependence of the partition coefficients upon the concentration in the sodium pool. The obtained empirical equations are consistent with Castleman's theoretical equations. The partition coefficients of cesium measured at five different points of mole concentration in the pool were almost consistent with the theoretical values. On the other hand, the measured partition coefficients of iodine increased with the increase in the concentration in the pool and this tendency was incompatible with the theoretical consideration. The reason of this discrepancy might be attributed to the formation of Na
I
in the cover gas.
Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Matsushita, Izuru*; Ida, Mizuho; Horiike, Hiroshi*; Kanemura, Takuji; Sugiura, Hirokazu*; Yagi, Juro*; Suzuki, Akihiro*; et al.
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
IFMIF is a neutron source aimed at producing an intense high energy neutron flux for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF started on July 2007. Regarding to the lithium (Li) target facility, design and construction of EVEDA Li Test Loop is a major activity and is in progress. The detail design was started at the early 2009. Fabrication of the loop was started at middle of 2009, and completion is planned at the end of Feb. 2011.
Li, Y.; Kaji, Yoshiyuki; Igarashi, Takahiro
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
Ito, Kei; Kunugi, Tomoaki*; Ohshima, Hiroyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
In the design study of large-sized sodium-cooled fast reactors in Japan (JSFR), the suppression of gas entrainment (GE) phenomena at a free surface in the reactor vessel is very important to establish an economically superior design. In this study, the unstructured adaptive mesh technique is developed for the numerical simulations of gas-liquid two-phase flows. The redistribution methods of two-phase flow variables are newly developed to satisfy the conservations of the variables, i.e. the volumes of gas and liquid phases, the location of interfaces and the momentum of each phase. This improved unstructured adaptive mesh technique for gas-liquid two-phase flows is validated by solving the well-known slotted disk revolution and dam-break problems.
Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Sogabe, Joji*; Deguchi, Yoshihiro*; Kikuchi, Shin
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
Sodium-water reaction (SWR) is a design basis accident of a sodium fast reactor (SFR). A breach of the heat transfer tube in a steam generator (SG) results in contact of liquid sodium with water. The purpose of the present paper is to delineate the mechanism and process of the SWR by a counter-flow diffusion flame experiment and a numerical simulation.
Ioka, Ikuo; Suzuki, Jun; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
An Extra High Purity austenitic stainless steel (EHP alloy) was developed with conducting the new multiple refined melting in order to suppress the harmful impurities less than 100 ppm. EHP alloy has great intergranular corrosion resistance. It is considered that intergranular corrosion becomes initiation of SCC. So, we try to apply EHP alloy to weld overlay materials to prevent from SCC. EHP alloy was melted by the new multiple refined method. The conventional weld metals were also prepared as comparisons. Specimens were machined from the welded metal of each material. Intergranular corrosion tests were performed in boiling 8 kmol/m HNO
solutions containing 1 kg/m
Cr(VI) ions. The intergranular corrosion of conventional weld metals was severer than those of EHP alloys. Crevice Beam bending tests to evaluate susceptibility of SCC were carried out in high temperature water of 561 K with saturated oxygen for 1000 h. Cracks and intergranular corrosion of conventional weld metals were much more than those of EHP alloys. It was confirmed that EHP alloy had excellent SCC resistance in comparison with conventional materials when EHP alloy was used as a weld metal.
Ide, Hiroshi; Kimura, Akihiro; Miura, Hiroshi; Nagao, Yoshiharu; Hori, Naohiko; Kaminaga, Masanori
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
Visual observation of inner side of a reactor pressure vessel (RPV) of JMTR was carried out using an underwater camera before the JMTR refurbishment work, because the RPV of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed were observed in a top closure. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the RPV was observed, and then the integrity of the RPV was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the RPV of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the RPV in future.
Igarashi, Takahiro; Aoyagi, Yoshiteru; Kaji, Yoshiyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 5 Pages, 2010/05
no abstracts in English