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Tanaka, Tadao; Nakayama, Shinichi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
no abstracts in English
Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500 kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port. Then another RH device receives and brings out the module by a pallet installed from outside the VV.
Igarashi, Takahiro; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
The two-dimensional intergranular stress corrosion cracking (IGSCC) growth model has been developed to simulate branching cracks of IGSCC. In the model, the IGSCC is grown using the "grain-scaled" factors such as the length and strength of grain boundary and so on. Especially, the corrosion of grain boundary and the influence of shear stress acting on the grain boundary are introduced in the model. Using the model, computer simulation of crack growth was carried out under several load conditions with changing the ratio of axial to shear stress against the grain boundary. As a result of the simulations, we found out that the cause of crack branching was the influence of shear stress against the grain boundary, and that the synergistic effect of shear stress and corrosion of grain boundary leads to the oblique crack growth.
Matsuba, Kenichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
Significant thermal stresses are loaded on the structures of sodium-cooled fast reactor (SFR) due to high temperature and large temperature gradients associated with employing sodium coolant. Therefore, it is important to monitor the temperature variation and related stress on the cooling system piping in order to assure structural integrity. Structural integrity monitoring can be enhanced by an optical fiber sensor, which is capable of continuous or dispersed distribution measurements of various physical properties such as radiation dose, temperature, strain, displacement and acceleration. In the experimental fast reactor Joyo, displacement and vibration measurements of the primary cooling system have been carried out using Fiber Bragg Grating (FBG) sensors to evaluate the durability and measurement accuracy of FBG sensors in a high 
ray environment. The data were successfully obtained with no significant signal loss up to an accumulated ray dose of approximately 4
10
Gy corresponding to 120EFPDs operation. Measured displacement of the piping support was nearly equal to the calculated thermal displacement. Measured vibration power spectra of the piping support were similar to those measured with a reference acceleration sensor. The measured results indicate that the FBG sensor is applicable for monitoring the displacement and vibration of fast reactor cooling system integrity in a high ray environment.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
The numerical analysis code, ACCORD, has modified to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core structural materials for calculating the heat conduction between the fuel channels and the core in the case of the coolant flow reduction test. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the coolant flow reduction test by tripping one or two out of three gas circulators. Finally, the pre-analytical result of the coolant flow reduction test by tripping all gas circulators is also discussed. The reactor power decreases to decay heat level from 30 MW due to the negative reactivity feedback effect. Although the reactor power becomes critical again about five hours later, the peak power value is merely 2 MW.
Terada, Atsuhiko; Hino, Ryutaro; Hirayama, Toshio; Nakajima, Norihiro; Sugiyama, Hitoshi*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
The Japan Atomic Energy Agency has been conducting study on thermochemical IS process for water splitting hydrogen production. Based on the test results and know-how obtained through the bench-scale test, a pilot test plant, which has a hydrogen production performance of 30 Nm
/h, is being designed conceptually as the next step of the IS process development. In design of the IS pilot plant, it is important to make chemical reactors compact with high performance from the viewpoint of plant cost reduction. A new hydraulic analytical code has been developed for optimizing mixing performance of multi-phase flow involving chemical reactions especially in the Bunsen reactor. Preliminary analytical results obtained with above mentioned code, especially flow patterns induced by swirling flow agreed well with that measured by water experiments, which showed vortex breakdown pattern in a simplified Bunsen reactor.
Nakayama, Okatsu; Ogawa, Hiroshi*; Kimura, Nobuyuki; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
Water experiment using an 1/10th scaled upper plenum model was carried out to investigate thermal stratification after a scram in a compact reactor, which has high velocity local flow in the upper plenum. The experiments showed that the rising speed of the stratification interface was dependent on Richardson number and the temperature gradient of the stratification interface was also influenced by the temperature difference and fluctuation. Furthermore, the temperature gradient could be reduced greatly by changing position of structure in the upper plenum.
Matsumura, Tatsuro; Takeshita, Kenji*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 5 Pages, 2007/04
We are developing a new MA/Ln separation process with TPEN derivatives for P&T technology. TPEN is a hexadentate ligand and a kind of podand type molecule and can encapsulate a metal ion. TPEN has good selectivity of Am(III) from Ln(III). However, there is a serious problem for the practical application. This is to the dissolution of a slight amount of TPEN to water. In this study, the hydrophobicity of TPEN is improved by introducing alkyl groups and the effect of the introduction of alkyl groups on the separation of Am(III) and Eu(III) is examined. We synthesized three derivatives successfully. The derivatives were examined both the extractability and selectivity of Am(III) and Eu(III). One of them, tpdben, showed good selectivity and the maximum separation factor, SF
, was 35 at pH 5.06. A hydrophobic derivative of TPEN that has potential of application to the MA/Ln separation process was synthesized successfully.
Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki; Iyoku, Tatsuo
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950
C in April, 2004 during the "rise-to-power tests" confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with the high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.
Sugo, Yumi; Sasaki, Yuji; Kimura, Takaumi; Sekine, Tsutomu*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 5 Pages, 2007/04
no abstracts in English
centuryUmeki, Hiroyuki; Naito, Morimasa; Makino, Hitoshi; Osawa, Hideaki; Nakano, Katsushi; Miyamoto, Yoichi; McKinley, I. G.*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
no abstracts in English
Kobayashi, Noboru; Onuki, Akira; Okubo, Tsutomu; Uchikawa, Sadao
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
A thermal-hydraulic design of the high-conversion (HC) type core of the innovative water reactor for flexible fuel cycle (FLWR) was constructed. HC-FLWR is required to proceed to the breeder type of FLWR with no change of any reactor systems. Although tightness of the fuel pin arrangement is significantly different between the two types of cores, the natural circulation cooling is adopted in both cores. TRAC analyses were performed under the condition that chimney length for natural circulation and the setting of the inlet orifice were common to the both types of cores. Form loss coefficients of lower tie-plate were differently set to control the natural circulation flow rate and feed water temperature were adjusted to realize preferable value of average void fraction of HC-FLWR core. The analyses showed that both types of the FLWR could be cooled by the same reactor system.
Shirasu, Noriko; Kuramoto, Kenichi; Nakano, Yoshihiro; Yamashita, Toshiyuki; Ogawa, Toru
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
To evaluate the irradiation behavior of the rock-like oxide fuel, irradiation experiments were carried out. Three fuels were prepared; a single phase fuel of yttria-stabilized zirconia containing UO
(U-YSZ) and two particle-dispersed fuels of U-YSZ particles in spinel or corundum matrix. These fuels were irradiated in JRR-3 for about 280 days. The burnups were about 11% FIMA. The fission gas release rate (FGR) was determined by puncture test and gas analysis. Corundum-based fuel showed extremely high FGR (88%). On the other hand, the U-YSZ single-phase fuel showed very low FGR (5%). Microstructure analyses for irradiated fuel pellets were carried out by ceramography and EPMA. The restructuring of fuel pellet was not observed in the spinel-based fuel irradiated below 1400 K. Significant appearance changes were not also observed for corundum-based fuel.
Kuji, Masayoshi; Sato, Toshinori; Mikake, Shinichiro; Hara, Masato; Minamide, Masashi; Sugihara, Kozo
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
The Mizunami Underground Research Laboratory (MIU) is being constructed. The MIU consists of two 1,000 m-deep shafts with several research galleries. The diameter of the shafts are 6.5 m and 4.5 m, respectively. Horizontal tunnels to connect the shafts are excavated at 100 m depth intervals. The Middle stage, at about 500 m depth, and the Main stage at about 1,000 m depth will be the main locations for scientific investigations. Current depths of shafts are 180 m and 191 m respectively, in November, 2006. During the construction, the quantity of water inflow into the shafts is increasing and disturbing the project progress. In order to reduce the quantity of water inflow, post-excavation grouting and pre-excavation grouting are planned. A test of post-excavation grouting was undertaken in the Ventilation shaft and the applicability of several techniques were evaluated.
Kitamura, Akihiro; Nakamichi, Shinya; Okada, Takashi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
Data on glovebox dismantling activities in the Glovebox Dismantling Facility (GDF) were analyzed to identify the work structure and the time consumed for each activity. As a result, we were able to categorize dismantling activities regarding time estimation point of view. The activities those of which variations are around 30% or less, were defined as "predictable activities", and activities those of which total time is small compare to the whole dismantling work were defined as "suppressible activities", and other as "unpredictable activities". In terms of these definition the time interval for unit activity were evaluated and found that almost all of the work can be predicted within 30% uncertainly.
Takeda, Tetsuaki; Ichimiya, Koichi*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 4 Pages, 2007/04
As for the development of the coupling technology between the HTGR and the hydrogen production system, JAEA have carried out the hydrogen production test with the steam reforming process by natural gas. In the HTGR hydrogen production system, disk type fins are attached on the outside surface of the catalyst tube and the tube is inserted into the guide tube to increase an amount of transferred heat in the present design of the steam reformer. However, we have to take the deterioration of the structure strength by attaching the fins and processing the tube surface into consideration with the design of the steam reformer. The objectives of this study are to develop a method for heat transfer enhancement using a porous material and to discuss the applicability of this method into the steam reformer of the nuclear hydrogen production system. An experiment has been performed using the simulated apparatus of the steam reformer to obtain the heat transfer and fluid flow characteristics.
Ohno, Shuji; Miyahara, Shinya; Kurata, Yuji
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
Evaporation tests were conducted to investigate fundamental liquid-to-gas transfer behavior of volatile radionuclides 
Po, Cs, and Te in a lead-bismuth eutectic (LBE) using the "transpiration" method, which has already contributed in a Na-cooled FBR study. Since both LBE test and Na test focus on the evaporation of the same nuclides, it is possible to compare the nuclides' volatility between in LBE and in Na. This paper describes first the reviewed evaporation characteristics of fission products in Na, next the evaporation test results of fission or activation products in LBE. The importance of investigating Po evaporation is demonstrated through the estimation of vapor amount in a cover gas region of an LBE-cooled system. Furthermore, comparison is made for the volatility of Cs and Te in two kinds of coolant Na and LBE based on the test results. The accumulated data can serve as significant database utilized in liquid-metal-cooled reactor accident analysis tools.
Saito, Junichi; Ara, Kuniaki; Sugiyama, Kenichiro*; Kitagawa, Hiroshi*; Oka, Nobuki*; Yoshioka, Naoki*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 5 Pages, 2007/04
Liquid sodium is used as the coolant of the fast breeder reactor (FBR), because of its high thermal conductivity and wide temperature range of liquid. However the chemical reactivity with water and oxygen of sodium is very high. So an innovative technology to control the reactivity is desired. The purpose of this study is to reduce the chemical reactivity of liquid sodium by dispersing the nanometer-size metallic particles into liquid sodium. Sub-themes of this study are nanoparticles production, evaluation of reaction control of liquid sodium, and feasibility study to FBR. In this paper, we describe the research program of them.
Onuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.
Sato, Haruo
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
no abstracts in English