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Suzuki, Mitsutoshi; Ihara, Hitoshi
Journal of Power and Energy Systems (Internet), 2(2), p.899 - 907, 2009/00
The large Pu throughput in an advanced reprocessing plant comes with the inevitable requirement of NMA. A large amount of sampling analysis result in a great cost to verify no undeclared use of Pu. In addition to NMA, C/S, PM, and Cu balance measurements have been used for SG systems. However, no quantitative formalism has been developed to date. Therefore, it is difficult to evaluate a cost-effective performance. In order to design an advanced SG system, JAEA has started to develop a SG system simulator. One core of the simulator for the NMA component is composed of a NRTA code that has been developed. In addition, a multivariate and multi-scale core has been developed. A concept for a multi-objective core is proposed for the SG formalism. Also, flow meter and NDA can be more broadly applied to the system. In future work, a virtual design model will be developed in the simulator.
Yamaguchi, Tetsuji; Nakayama, Shinichi; Vandergraaf, T. T.*; Drew, D. J.*; Vilks, P.*
Journal of Power and Energy Systems (Internet), 2(1), p.186 - 197, 2008/00
In safety assessments of the geologic disposal of high-level radioactive wastes, the possibility that long-lived radionuclides may be leached from the wastes and may subsequently be transported through surrounding rock masses must be considered. It is therefore necessary to understand the transport of radionuclides through water-bearing fractures in rocks surrounding the repository. For this purpose, radionuclide migration experiments in quarried blocks of granite under in-situ conditions at the 240-m level in AECL's Underground Research Laboratory (URL) were performed under a five-year cooperative research program between Japan Atomic Energy Research Institute (JAERI, reorganized to Japan Atomic Energy Agency, JAEA) and Atomic Energy of Canada Ltd. (AECL). Migration experiments with Br, synthetic colloids, 
H, 
Sr, 
Tc, 
Np and 
Pu, and post-experimental alpha and 
scanning of the fracture surfaces were performed using 1 m granite blocks, each containing a single fracture, excavated from a water-bearing fracture zone. The transport of the radionuclides was affected by macroscopic mechanical dispersion, matrix diffusion and element specific sorption on fracture surfaces. Colloid transport exhibited a complicated process that may include sedimentation and diffusion into stagnant zones.
Sato, Haruo
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
no abstracts in English
Tanaka, Tadao; Nakayama, Shinichi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 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
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.
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.
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.
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.
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.
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
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.
