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Aoyama, Yoshio; Miyamoto, Yasuaki; Yamaguchi, Hiromi; Sano, Akira*; Naito, Susumu*; Sumida, Akio*; Izumi, Mikio*; Maekawa, Tatsuyuki*; Sato, Mitsuyoshi*; Nambu, Kenichi*; et al.
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
no abstracts in English
Hirata, Yosuke*; Nakahara, Katsuhiko*; Sano, Akira*; Sato, Mitsuyoshi*; Aoyama, Yoshio; Miyamoto, Yasuaki; Yamaguchi, Hiromi; Nambu, Kenichi*; Takahashi, Hiroyuki*; Oda, Akinori*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
no abstracts in English
Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Suzuki, Toru; Kamiyama, Kenji; Morita, Koji*; Maschek, W.*; Pigny, S.*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
To simulate complex phenomena during core disruptive accidents in sodium-cooled fast reactors, JAEA has been developing the SIMMER-III code,which is two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. Recently, the three-dimensional code SIMMER-IV is also developed with the same physical model as SIMMER-III. In the present paper, the models and methods of SIMMER-III/IV are briefly reviewed with highlighting the recent improvements. The major achievements of the code assessment program are then described, followed by presentation of practical applications. A three-dimensional calculation with SIMMER-IV are also shown to indicate more realistic accident scenario. In addition, this calculation result show the disrupted core state for investigating the post-accident material relocation and heat removal phase.
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.
Tamura, Takeshi*; Soga, Kazuo*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Suda, Kazunori
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
To evaluate the phenomena for the secondary heat transfer tube failure due to overheating by sodium-water reaction in steam generator of liquid sodium cooled fast breeder reactor, a sophisticated computer code SERAPHIM has been developed by JAEA. As the first step to verify the adequacy of SERAPHIM code, a visualization experiment of Ar gas jet impinging a single rod with 20mm in diameter immersed in a water pool was performed in our previous study. In this paper, we measured the void fraction around a single rod in the water pool as a basic experiment using the apparatus capable of doing sodium pool experiment to investigate the flow pattern and the water entrainment around a single rod. The result of the void fraction reflected the result of the heat transfer experiment that had been reported before was obtained, and a certain prospect of the measurement of the void fraction in sodium pool was obtained.
Sato, Hiroyuki; Ohashi, Hirofumi; Sakaba, Nariaki; Nishihara, Tetsuo; Kunitomi, Kazuhiko
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
Japan Atomic Energy Agency (JAEA) has been conducting R&D on the hydrogen production system to be coupled with the high-temperature gas-cooled reactors (HTGRs). Thermochemical water-splitting iodine-sulphur process (IS process) is a progressive candidate for its hydrogen production system. Since the reactor needs to keep its operation during abnormal events caused by the IS process in the hydrogen production system coupled with HTTR (HTTR-IS system), countermeasure for the abnormal accidents caused by IS process should be established. In this study, assumed abnormal accidents caused by IS process was extracted and dynamic behaviour of the HTTR-IS system during the abnormal condition was calculated by the newly developed dynamic simulation code based on the RELAP5 code. It was confirmed that the cooling system using steam generator with air cooler had superb functionality to mitigate the influence of abnormal events caused by the IS process.
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.
Iyoku, Tatsuo; Nojiri, Naoki; Tochio, Daisuke; Mizushima, Toshihiko; Tachibana, Yukio; Fujimoto, Nozomu
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
A HTGR is particularly attractive because of its capability of producing high temperature helium gas and its inherent safety characteristics. Hence, the HTTR wasconstructed at the Oarai Research Establishment of the Japan Atomic Energy Agency. The HTTR achieved the full power of 30MW and reactor outlet coolant temperature of about 850
C on December 7, 2001. After several operation cycles, the HTTR achieved the reactor outlet coolant temperature of 950C on April 19, 2004. It is the highest coolant temperature outside reactor pressure vessel in the world. Extensive tests are planned in the HTTR and a process heat application system will be coupled to the HTTR, where hydrogen will be produced directly from the nuclear energy.
Nobuto, Jun*; Kobayashi, Shinji*; Nishigaki, Makoto*; Mikake, Shinichiro; Sato, Toshinori
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
Authors are conducting research on construction technologies such as water sealing measures as a scientific and technological basis for geological disposal. In this paper, the application patterns of grouting technology to excavation work in deep underground are selected first, and potential grouting material for each pattern is summarized. Next, the method to achieve high level water tightness in crystalline rock with suspension grout like cement and bentonite slurry is studied. Finally, focusing on pre-excavation grouting for horizontal tunnels, the effective injection method is studied.
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.
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