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Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/07
During the investigation of an incident in Joyo, in-vessel observations using a Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A was photographed using a VC fixed to the rotating-plug for making observations of the top of S/As and IVS. A resolution was estimated to be approximately 1mm. In order to observe the bottom face of the UCS, a remote handling device equipped with RRFs was specifically developed for Joyo with a tip that can be bent into an L-shape and inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. The sodium flow regulating grid of 0.8mm in thickness could be discriminated. These experiments provided valuable insights for use in further improving and verifying in-vessel observation techniques in sodium cooled fast reactors.
Shibamoto, Yasuteru; Maruyama, Yu; Nakamura, Hideo
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/07
A series of experiments was performed for rewetting phenomena on dried-out heated surfaces under post-BT (Boiling Transition) conditions with high-pressure and high-water flow rate simulating anticipated operational occurrences of a BWR. An analytical model for rewetting velocity, defined by the propagation velocity of quench front, has been developed on the basis of the experimental results. The experiment was conducted within the ranges of the flow rate and the pressures covering an actual BWR plant conditions. The rewetting for the post-BT conditions is characterized by faster propagation of the quench front than that for reflood phase conditions during a postulated large-break loss-of-coolant accident. In order to provide an explanation of this characteristic, the present analytical model took the effect of precursory cooling into account by modifying the existing correlation of Sun-Dix-Tien (1975) which is based on the one-dimensional analysis in a flow direction during the reflood phase. The present model demonstrates that the precursory cooling can significantly increase the rewetting velocity by more than an order of magnitude. Appling the experimental correlation developed in the separately conducted experiment into the heat transfer coefficient in the present model at a wet and a dry region with precursory cooling, our experimental data of the rewetting velocity as well as the wall temperature profiles for the variable flow rates are successfully predicted. It is found that the effect of precursory cooling is indispensable to explain the considerably high rewetting velocity under the large flow rate condition due probably to the significant droplets cooling contribution.
Shimada, Taro; Oshima, Soichiro; Sukegawa, Takenori
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/07
A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning of the plant, and also evaluates the public dose at accidental situations including fire and explosion. The DecDose is expected to contribute to utilities in formulating rational dismantling plans and to the safety authority in estimating conservativeness in safety assessment of licensing application or risk-based regulatory criteria.
Liu, W.; Takase, Kazuyuki
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 6 Pages, 2009/07
A measurement system for surface temperature and surface heat flux was developed to study heat transfer mechanism in boiling process. The system was consisted by two parts: (1) inner block temperatures were measured using micro-thermocouples set at two layers inside heating block; (2) with using the measured temperatures, inverse heat transfer analysis was performed to get surface heat flux and surface temperature. For the inner block temperature measurement, special T-type micro thermocouples with a common positive pole were developed. The developed system was used to research the change of surface heat flux and surface temperature in a boiling process. The experiments showed the developed special T-type micro thermocouples could trace temperature change in boiling process successfully. Increase in surface heat flux with the generation of big bubble was derived successfully.
Xu, Y.*; Tagawa, Akihiro; Fujiki, Kazunari*; Ueda, Masashi; Yamashita, Takuya
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 9 Pages, 2009/07
In-service inspection is carried out to confirm the integrity of the main components of "MONJU". The ambient temperature during the inspection is 200 
C and the irradiation field is 10 Gy/hr. A Periodic Permanent Magnet (PPM) structure has been used for electromagnetic acoustic transducer (EMAT). The PPM structure is the structure of putting in order alternately the S pole and N pole laterally magnetized magnets. A Halbach magnet structure is the structure which sandwiched the longitudinally magnetized magnets between each magnet of PPM structure. Thereby, magnetic flux density is improved 1.4 times. Moreover, it also checked aiming at 4 times as many improvement in detection sensitivity as this by a new signal processing method. The high temperature characteristic and high temperature durability of Halbach type EMAT in 200 hours which corresponds the twice of inspection time under the 200 C inspection environment are reported.
Takahashi, Kenji; Shiina, Akira; Onizawa, Takahiro; Ibaki, Shoji; Yamaguchi, Toshihiko; Tagawa, Akihiro
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 9 Pages, 2009/07
Japanese prototype FBR Monju will restart its operation for the first time since the sodium leak accident in 1995. Concerning heat transfer tubes of the steam generator (SG) units, presumable deteriorations were studied and the integrity is confirmed when they have no lack of thickness by corrosion after the long-term suspension. It was impossible to test the tubes directly, then three tests were applied, which were ECT, VT (visual test) and leak test. ECT is to check for the lack of thickness caused by regional corrosion. VT checked for the global corrosion and the bottom of the tube. Leak test checked for pin-holes (penetrating holes). Total evaluation proved no significant lack of thickness and pin-holes in the tube.
Kim, G.; Suzuki, Yoshio; Teshima, Naoya
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/07
Ito, Kei; Kunugi, Tomoaki*; Ohshima, Hiroyuki
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 9 Pages, 2009/07
In a design study of the large-sized sodium-cooled fast reactors in Japan, one key issue to establish an economically superior design is the suppression of a gas entrainment phenomenon (GE). Therefore, we are developing a high-precision CFD method to evaluate the GE phenomenon accurately. In this study, as one part of the development, a two-dimensional unstructured adaptive mesh technique is developed and verified. In the two-dimensional unstructured adaptive mesh technique, each cell is isotropically subdivided to reduce distortions of cells. In addition, a connection cell is formed to eliminate the edge incompatibility between refined and non-refined cells. Finally, the present unstructured adaptive mesh technique is verified by solving well-known driven cavity problem and as a result, the present unstructured adaptive mesh technique succeeds in providing a high-precision solution with less number of cells than the structured mesh.
Nishida, Akemi
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 7 Pages, 2009/07
It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.
Sako, Hiroyuki
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/07
J-PARC is a multi-purpose research facility for materials and life sciences, nuclear and particle physics, and nuclear engineering. The accelerator complex consists of a 400-MeV linac, a 3-GeV Rapid Cycling Synchrotron (RCS), and a 50-GeV Main Ring synchrotron (MR). To provide MW-class beams at 3 GeV and several 10 GeV while localizing and suppressing beam loss for hands-on maintenance, challenging technologies such as ceramic ducts without water cooling and high field RF system with Magnetic Alloy in RCS, and compact electromagnet Drift-Tube Quadrupoles in the linac were developed. The linac provides stable beam with short down time. RCS recorded 0.21 MW in Sep. 2008 with localized loss at the collimators. The linac will be upgraded to 400 MeV with Annular Coupled Structure linac. In the second construction phase, upgrade of the linac with 600-MeV Super-Conducting Linac for Accelerator-Driven nuclear transmutation System and upgrade of MR to 50 GeV are planned.
Nishizaki, Masanori*; Tsuruoka, Hokuto*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 6 Pages, 2009/07
The secondary tube failure may occur due to overheating by sodium-water reaction in LMFBR steam generator. It is very important to understand the void fraction distribution in sodium pool to evaluate the overheating tube rupture. In the present study, the Ar jet of 17.3 m/s to 129.8 m/s was injected from nozzle of 3.5 mm diameter in sodium pool with 443 K and 293 K. The authors measured the void fraction without chemical reaction along the jet-center axis. As the result, the void fraction increased when the distance from the nozzle decreased. The void fraction did not change when the distance from the nozzle was blow or equal to about 1.0 mm. The void fraction in sodium was lower the that in water, it is suggested that this trend reflects the fact the surface tension of sodium is higher than that of water.
Zhang, W.*; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/07
no abstracts in English
Kitagaki, Toru; Tasaka, Masayuki; Higuchi, Hidetoshi; Koizumi, Kenji; Hirano, Hiroyasu; Washiya, Tadahiro; Kobayashi, Tsuguyuki*
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 5 Pages, 2009/06
JAEA has been developing a reliable disassembly system for FBR fuel reprocessing as a part of Fast Reactor Cycle Technology Development (FaCT). We fabricated the disassembly system testing machine. In this paper, we described design of it.
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06
This paper describes experimental analyses using SIMMER-III, which were precedently carried out for the integral verification of the COMPASS code. Two topics of key phenomena in CDAs were presented in this paper: fuel-pin disruption; and low-energy disrupted core motion. To analyze the fuel-pin disruption behavior, the CABRI-EFM1 and the CABRI-E7 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall fuel-pin disruption and dispersion behavior, which was characterized by a thermal pin-failure mode, observed in the CABRI-EFM1 experiment. A mechanical pin-failure mode observed in the CABRI-E7 experiment was also reasonably simulated. Under a low-energy disrupted core condition, significantly reduced melt penetration length was obtained in the THEFIS out-of-pile experiments. SIMMER-III well simulated the melt freezing and blockage behavior observed in the experiment.
Ioka, Ikuo; Suzuki, Jun; Motooka, Takafumi; Kiuchi, Kiyoshi; Nakayama, Jumpei
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 5 Pages, 2009/06
An intergranular corrosion is an important degradation mechanism of austenitic stainless steels for use in a nuclear fuel reprocessing plant. The intergranular corrosion is caused by the segregation of impurities to grain boundaries. An extra high purity austenitic stainless steel (EHP
) was developed with conducting the new multiple refining to suppress the impurities less than 100 ppm. The intergranular corrosion behavior of EHP alloys with various impurities was examined in boiling nitric acid solution. The intergranular corrosion was observed in impurity doped EHP alloys, although no intergranular corrosion was observed in EHP alloy. From the obtained results, an empirical equation between susceptibility of intergranular corrosion and impurities was established by means of the regression analysis. The degree of influence of the impurities on intergranular corrosion was shown.
Furukawa, Tomohiro; Inagaki, Yoshiyuki; Aritomi, Masanori*
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 6 Pages, 2009/06
A key problem in the application of a supercritical carbon dioxide (CO) turbine cycle to a fast breeder reactor is the corrosion of structural materials by supercritical CO at high temperature. In this study, corrosion tests on the candidate materials, high-chromium martensitic steel (12Cr-steel) and FBR grade type 316 stainless steel (316FR), were performed for up to approximately 2000h at 400-600 C in supercritical CO pressurized at 20 MPa.
Ohira, Hiroaki; Takamatsu, Misao
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/06
In the incident of the experimental fast reactor Joyo on June 2007, little amount of metal powder was estimated to be produced by the contact between the bottom of the upper core structure and the test section of the material testing rig with temperature control. Up to now, no foreign materials were detected by the in-vessel observations. However, a small amount of metal powder which could be placed into the fuel pin bundles was assumed in the present study. Preliminary safety evaluation of the wire-wrapped fuel pin bundles of Joyo with local blockage were performed in the rated power operational condition, an anticipated transient during operation and an accident. A single-phase transient subchannel analysis code ASFRE, which was verified by various local blockage experiments in France and in Japan, was applied to the present evaluations. From these results, it was concluded a small amount of metal powder assumed to be produced by the incident would not affect to the safety operations in rated power conditions, in anticipated transients and in accidents.
Sotsu, Masutake; Kurisaka, Kenichi
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 9 Pages, 2009/06
MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops which can supply 280 MW of electricity. Limiting conditions of operation (LCO) defined in the safety regulations in MONJU given the allowed outage time (AOT) are evaluated using a PSA technique. The result indicates the possibility of limit extension and some prospects that we should examine.
Yamamoto, Yuichi*; Hirano, Etsujo*; Oue, Masaya*; Shimizu, Sensuke*; Shirakawa, Noriyuki*; Koshizuka, Seiichi*; Morita, Koji*; Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06
The COMPASS code is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of MPS method. In FY2006 and 2007, development of the basic functions of COMPASS was completed and fundamental verification calculations were carried out. In FY2007, the integrated verification program using available experimental data for key phenomena in CDAs was also started. In this paper, we show the basic verification calculations for the phase change model of COMPASS and the results of experimental analyses, together with the outline of the formulation of MPS method and the conceptual design of the COMPASS code.
Uehara, Yasushi*; Shirakawa, Noriyuki*; Naito, Masanori*; Okada, Hidetoshi*; Yamano, Hidemasa; Tobita, Yoshiharu; Yamamoto, Yuichi*; Koshizuka, Seiichi*
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06
A mesoscopic approach with the COMPASS code is expected to advance the understanding of key phenomena during event progression in core disruptive accidents. In this paper, the overall analysis of SCARABEE-BE+3 test with the SIMMER-III is described as well as the simulation with COMPASS, focusing on the duct wall failure in a small temporal and spatial window cut from the SIMMER-III analysis results.