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Muramatsu, Ken; Liu, Q.; Uchiyama, Tomoaki*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 9 Pages, 2008/01
Aiming at proposing effective applications of seismic PSA for risk management of nuclear facilities, a preliminary study was conducted for a multi-unit site to examine core damage frequency (CDF) and core damage sequences with consideration of correlation of component failures. Besides, the effectiveness of an accident management measure, namely, cross connection of emergency diesel generators (EDGs) between adjacent units was also examined. Twin BWR-5 units of the same design were hypothesized to be located at the same site and the CDF as well as the accident sequences of this site was analyzed by using SECOM2. The results showed that the calculated CDF depends on the assumptions on component correlations and when the rules for assigning correlation coefficients defined in NUREG-1150 program was adopted, the CDF of a single unit, the CDF of this site and the frequency of simultaneous core damage of both units increased. Further, it might be possible that simultaneous core damage of both units was caused by pairs of different accident sequences. When cross-connection of EDGs between two units was available, the CDF of a single unit, the CDF of this site as well as the frequency of simultaneous core damage of both units decreased, which indicated that cross-connection of EDGs can be an effective measure for reducing seismic risk.
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
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.
Soga, Tomonori; Sekine, Takashi; Wootan, D. W.; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
The MA-containing mixed oxide fuel irradiation program is being conducted using the experimental fast reactor Joyo of the JAEA. Two irradiation tests were conducted in the JOYO MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing Am (Am-MOX) or MOX fuel containing Am and Np (Np/Am-MOX). The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, an Am-MOX pin and a Np/Am-MOX pin in the test subassembly were replaced with dummy pins. Remaining four fuel pins were re-irradiated in Joyo for 24-hours in August 2006 to obtain MA re-distribution data. Linear heat rates for each pin were calculated using MCNP and then adjusted using reaction rates of neutron dosimeter measured in the MK-III core. Post irradiation examination of these pins to confirm the fuel melting and the re-distribution are underway.
Ito, Chikara; Kagota, Eiichi; Ishida, Koichi; Kitamura, Ryoichi; Aoyama, Takafumi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 9 Pages, 2007/04
no abstracts in English
Kuno, Yusuke
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
Since many clandestine nuclear activities in the Middle East, the Korean peninsula and other areas of the world have been disclosed during the last 15 years, a series of the counter-measures have been proposed and taken. The Destructive Analysis (DA) for environmental sampling for safeguards (ESS), based on the Additional Protocol to the NPT Safeguards Agreement has played very important role since 1990s. Among the verification tools for the comprehensive Safeguards agreement, DA for nuclear material accountancy and its verification is extremely important for drawing quantitative Safeguards conclusions. Nuclear accountancy and verification based on the DA with a state-of-the-art determination technique providing highest possible measurement accuracy is the fundamental and essential technology, without which Safeguards cannot be concluded. This paper describes the role of DA for both environmental sampling and verification of nuclear material accountancy.
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.
Sahara, Fumihiro*; Murakami, Takeshi*; Mihara, Morihiro; Oi, Takao
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
An analysis system for the long-term mechanical behavior of barrier materials (MACBECE: Mechanical Analysis system considering Chemical transitions of BEntonite-based and CEment-based materials) was developed in order to improve the reliability of the evaluation of the hydraulic field which is one of the important environmental conditions in the safety assessment of the TRU waste disposal. MACBECE is the system that calculates the deformation of barrier materials using their chemical property changes as inputs, and subsequently calculates their hydraulic conductivity taking both their chemical property changes and deformation into consideration. By using MACBECE, the long-term deformation and the transition of hydraulic field for the round-type disposal cavities were evaluated, assuming a set of chemical evolution data as input. Based on the analysis result, it is considered that the influence of the long-term deformation of the barrier materials on the nuclide migration is not necessarily significant.
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.
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.
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
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.
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.
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.
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.
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
