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Sato, Yuki; Terasaka, Yuta; Miyamura, Hiroko; Kaburagi, Masaaki; Tanifuji, Yuta; Kawabata, Kuniaki; Torii, Tatsuo
Reactor Dosimetry; 16th International Symposium on Reactor Dosimetry (ISRD-16) (ASTM STP 1608), p.428 - 436, 2018/11
Times Cited Count：0 Percentile：0.05(Nuclear Science & Technology)Kaburagi, Masaaki; Sato, Yuki; Yoshihara, Yuri*; Shimazoe, Kenji*; Takahashi, Hiroyuki*; Torii, Tatsuo
Reactor Dosimetry; 16th International Symposium on Reactor Dosimetry (ISRD-16) (ASTM STP 1608), p.405 - 414, 2018/11
Times Cited Count：0 Percentile：0.05(Nuclear Science & Technology)Takeuchi, Tomoaki; Ueno, Shunji; Komanome, Hirohisa*; Otsuka, Noriaki; Shibata, Hiroshi; Kimura, Nobuaki; Matsui, Yoshinori; Tsuchiya, Kunihiko; Araki, Masanori
Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 7 Pages, 2013/10
During the station blackout situation at the Fukushima Dai-ichi (1F) Nuclear Power Plant, conventional in-pile instrumentation systems did not work sufficiently, resulting in the progress of the severe accident. In June 2011, the Japanese government referred to "Enhancement of instrumentation to identify the status of the reactors and PCVs" as a lesson of the accident at the 1F NPP, in the report of Japanese government to the IAEA ministerial conference in accordance with such situation, we started from 2012 a research and development which corresponds to the provisions so as to monitor the NPPs situations during a severe accident. In this research and development, we have been building of technical bases of a radiation-resistant high-definition and high-sensitivity monitoring camera, a wireless transmission system, and radiation- and heat-resistant signal line. The objective and latest progress situations of the R&D including the results of the characteristic experiments will be introduced in this symposium.
Kakei, Sadanori*; Kimura, Akihiro; Niizeki, Tomotake*; Ishida, Takuya; Nishikata, Kaori; Kurosawa, Makoto; Yoshinaga, Hideo*; Hasegawa, Yoshio*; Tsuchiya, Kunihiko
Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 7 Pages, 2013/10
The Japan Materials Testing Reactor (JMTR) is expected to contribute to the expansion of industrial utilization, such as the domestic production of 
Mo for the medical diagnosis medicine 
Tc. Production by the (n, 
) method is proposed as domestic Mo production in JMTR because of the low amount of radioactive wastes and the easy Mo/Tc production process. Molybdenum oxide (MoO
) pellets, poly zirconium compounds (PZC) and poly titanium compounds (PTC) are used as the irradiation target and generator for the production of Mo/Tc by the (n, ) method. However, it is necessary to use the enriched 
MoO, which is very expensive, to increase the specific activity of Mo. Additionally, a large amount of used PZC and PTC is generated after the decay of Mo. Therefore, this recycling technology of used PZC/PTC has been developed to recover molybdenum (Mo) as an effective use of resources and a reduction of radioactive wastes. The total Mo recovery rate of this process was 95.8%. From the results of the hot experiments, we could demonstrate that the recovery of MoO and the recycling of PZC are possible. In the future, the equipment of recovering Mo will be installed in JMTR-Hot Cell, and this recycling process will be able to contribute to the reduction of production costs of Tc and the reduction of radioactive wastes.
Mo-Tc production process by (n,) reaction with irradiated high-density MoO pelletsTsuchiya, Kunihiko; Nishikata, Kaori; Tanase, Masakazu*; Shiina, Takayuki*; Ota, Akio*; Kobayashi, Masaaki*; Yamamoto, Asaki*; Morikawa, Yasumasa*; Takeuchi, Nobuhiro*; Kaminaga, Masanori; et al.
Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 9 Pages, 2013/10
no abstracts in English
Dorn, C. K.*; Tsuchiya, Kunihiko; Takemoto, Noriyuki; Ito, Masayasu; Hori, Junichi*; Chekushina, L.*; Hatano, Yuji*; Chakrov, P.*; Kawamura, Hiroshi
Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 9 Pages, 2013/10
no abstracts in English
Tc production from (n,)Mo based on solvent extraction and column chromatographyKimura, Akihiro; Awaludin, R.*; Shiina, Takayuki*; Tanase, Masakazu*; Kawauchi, Yukimasa*; Gunawan, A. H.*; Lubis, H.*; Sriyono*; Ota, Akio*; Genka, Tsuguo; et al.
Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 7 Pages, 2013/10
JP, 2011-173260
Patent publication (In Japanese)
This research is development of Tc production. Tc is generated by decay of Mo. The supply of Mo in Japan depends entirely on the import from foreign countries. Thus, it is needed to supply Mo stably by the domestic manufacturing. A practical production of Tc was tried by the method with 1 Ci of Mo produced in MPR-30. The results showed that the recovery yields were approximately 70%. The concentration of the product obtained was estimated to be corresponding to about 30 GBq (800 mCi)/ml when 150g of MoO was irradiated for 5 days in MPR-30. Impurity of Mo was less than 4.4
10
%, which was lower than that of Japanese tentative regulation criteria. The radiochemical purity was higher than 99.8% that cleared the tentative regulation (95%) of Japan.
Takase, Kazuyuki; Misawa, Takeharu; Yoshida, Hiroyuki; Mori, Hideo*
Proceedings of 6th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-6) (CD-ROM), 9 Pages, 2013/03
Numerical analyses of crossing flows between two parallel circular channels were conducted for a specific geometry that simply modeled subchannels in a fuel bundle of a supercritical water reactor. The two parallel circular channels were connected by a rectangular channel in the axial direction. Crossing flow occurred in the rectangular channel and was caused by differences in temperatures of fluids flowing in the two channels. The working fluid was supercritical Freon. The SST turbulence model was chosen for precisely calculating the boundary layers of temperature and velocity near the channel walls. From the analytical results, relations between crossing flow and fluid temperature were clarified quantitatively.
Eguchi, Yuzuru*; Murakami, Takahiro*; Ohshima, Hiroyuki; Yamano, Hidemasa; Kotake, Shoji
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
The unsteady turbulent flow in a short-elbow pipe to be employed in a Japanese sodium-cooled fast breeder reactor was computed to examine the fundamental features of the flow, especially, pressure fluctuation to cause unsteady fluid force on the pipe. An FEM-based large-eddy simulation code, named SMART-fem, was used for the computation. The results at Re=3.210
and 1.210
show that two separation regions exist on the inner urvature of the elbow around 45-degree (middle of elbow) and 90-degree (end of elbow) positions. The statistical quantities of pressure fluctuation such as deviation, skewness and flatness were computed and analyzed, showing that there exist two symmetric regions of significant pressure fluctuation on the wall of inner curvature of the elbow. It has turned out that the pressure loss coefficient of the elbow pipe agrees well among the computation, experiment and authoritative reference data.
Iwamoto, Yukiharu*; Yasuda, Kazunori*; Sogo, Motosuke*; Yamano, Hidemasa; Kotake, Shoji
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Pressure measurement, laser Doppler velocimetry (LDV) and flow visualization were carried out using the 1/10-scale model of a hot leg piping installed in a Japanese sodium-cooled fast breeder reactor. LDV measurement with Reynolds number of 50000 showed the following results: (1) A flow separation was confirmed in the region between 45 degrees from the elbow inlet and 0.3 times of pipe diameter downstream of the elbow. (2) There appeared two kinds of fluctuations in the present study. In the case of Reynolds number of 320000, it was found that the height of the flow separation downstream of the elbow became smaller, since the inertia of the flow became superior to the inverse pressure gradient.
Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 4 Pages, 2008/11
In order to confirm thermal safety of the supercritical-water-cooled reactor (SCWR), it is important to assess the thermal hydraulics in rod bundles of the core. In the present study, the three-dimensional two-fluid model analysis code ACE-3D, which has been developed in JAEA for the two-phase flow thermal-hydraulics of light water reactors, was improved to handle the thermal hydraulic properties of water at supercritical region. Heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which simulates core flow around a fuel rod, were analyzed with the ACE-3D to assess the prediction performance of the code. As a result, it was confirmed that the calculated wall surface temperature agreed with the measured results and the code is applicable to prediction of heat transfer of supercritical water in the system that simulates the SCWR core. To improve prediction accuracy for heat transfer deterioration is a subject for future study.
Hirota, Kazuo*; Ishitani, Yoshihide*; Nishida, Keigo*; Sago, Hiromi*; Xu, Y.*; Yamano, Hidemasa; Nakanishi, Shigeyuki; Kotake, Shoji
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
CFD simulation using the Reynolds stress model was performed to evaluate turbulence-induced forces on the piping. The turbulence energy with the CFD simulation was compared with pressure fluctuation distributions obtained by the test with a 1/3 scale elbow simulating the JSFR hot-leg piping. The profile of turbulence energy was good agreement with that of the pressure fluctuation. The magnitude of pressure fluctuation can also be estimated from calculated turbulence energy multiplied by a certain coefficient. In the vibration analysis, the power spectrum density (PSD) of the pressure fluctuation was derived from the measured normalized PSD multiplied by the coefficient. The vibration analysis method was proposed based on the PSDs derived by the above procedure and correlation lengths. The analysis results of vibration response showed good agreement with the flow-induced-vibration test results, thereby it can be said that the vibration analysis method developed in this study is valid.
Aizawa, Kosuke; Nakanishi, Shigeyuki; Yamano, Hidemasa; Kotake, Shoji; Hayakawa, Satoshi*; Watanabe, Osamu*; Fujimata, Kazuhiro*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
To evaluate the flow-induced vibration in the actual-sized pipings of JSFR, computer simulation is necessary. In this study, as the first step, sensitivity analysis of turbulence flow models for unsteady short-elbow pipe flow has been carried out with the STAR-CD thermal-hydraulic simulation code. Through the sensibility analysis, the objective of this study is to propose the best analysis models which can reproduce the unsteady characteristics obtained in the 1/3-scale test results with 9.2 m/s of main flow. In this study, to take into account anisotropic characteristics of turbulence, two turbulent flow models were used: large eddy simulation (LES) and Reynolds stress model (RSM). The both validated simulations have reproduced flow separation region and periodic vortex shedding. The simulation results with both models were compared with power spectrum densities of pressure fluctuations which were used in the pipe vibration evaluation. Only the RSM simulation with the best combination has reproduced the pressure-fluctuation power spectrum densities, which were characterized by a peak frequency of 10 Hz in the 1/3 test with 9.2 m/s.
Yoshida, Hiroyuki; Nakatsuka, Toru; Suzuki, Takayuki*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Kamide, Hideki; Aizawa, Kosuke; Oshima, Jun*; Nakayama, Okatsu*; Kasahara, Naoto
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
Development of advanced loop type sodium cooled fast reactor is under going. An upper internal structure (UIS) has a radial slit to reduce the reactor vessel diameter. This UIS slit allows a high velocity from the core fuel subassemblies and influences the gas entrainment in the reactor vessel and also the delayed neutron precursor sampling for a failed fuel detection and location system. Then flow visualization and velocity measurements were carried out in an 1/10 scale water test model. The velocity measurement using particle image velocimetry showed that velocity in the slit region was accelerated at the heights of the UIS horizontal plates and kept higher value at the middle height of the upper plenum. Numerical simulation using a commercial CFD code was also carried out for this complex geometry of UIS to know adequate simulation method. The comparisons of velocity profiles in the UIS between the experiment and analysis showed good agreements.
Ohno, Shuji; Oki, Hiroshi*; Sugahara, Akihiro*; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
Validation study of numerical simulation method is in progress for thermal stratification phenomena in a reactor vessel upper plenum of advanced sodium-cooled fast reactors. This paper describes the current status of the study using two kinds of thermal stratification experiments and commercial CFD codes STAR-CD, FLUENT, and an in-house code AQUA.
Takeda, Takeshi; Asaka, Hideaki*; Watanabe, Tadashi; Nakamura, Hideo
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
Two LSTF experiments were conducted for OECD/NEA ROSA Project simulating PWR 0.5% cold leg small break LOCA. Steam generator (SG) secondary-side depressurization was performed by fully opening the relief valves at 10 minutes after safety injection signal with or without non-condensable gas (air) inflow from accumulator tanks with total failure of high pressure injection system. Further assumptions were made to conduct enhanced SG depressurization by fully opening the safety valves when the primary pressure decreased to 2 MPa and no actuation of low pressure injection system, both to well observe natural circulation (NC) phenomena at low pressures. The primary depressurization rate decreased when non-condensable gas started to enter primary loops because of degradation in the condensation heat transfer in SG U-tubes, while two-phase flow NC has continued even after non-condensable gas inflow. Asymmetric NC behaviors appeared between two loops due probably to different number of forward flow SG U-tubes which would have been under influences of non-condensable gas. Post-test analyses by using JAEA-modified RELAP5/MOD3.2.1.2 code indicated that the code has remaining problems in proper prediction of primary loop flow rate and SG U-tube liquid level behaviors especially after non-condensable gas inflow. The improvement of the condensation heat transfer model under non-condensable gas mixture condition and the SG U-tube model may be necessary for correct analysis of the LSTF SG depressurization transients.
Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Gas entrainment (GE) at free surface is one of the significant issues for the design of Japanese Sodium Cooled Fast Reactor. In the previous study, authors confirmed that GE did not occur at the rated operating mode in the reactor. In the present study, a water experiment to simulate the startup operation of reactor was performed in a large-scaled partial model of the reactor upper plenum in the reactor. The onset condition of GE was observed by the visualization of free surface. Vertical velocity distribution was also measured in order to qualify the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of reactor.
Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V). An upper inner structure (UIS) has a slit in radial direction for fuel handling. A water experiment using an 1/10 scale model was carried out. Steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at a dipped plate for a fuel handling machine, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 13% smaller than that in the case of the higher plug position.
Ito, Kei; Eguchi, Yuzuru*; Monji, Hideaki*; Ohshima, Hiroyuki; Uchibori, Akihiro; Xu, Y.*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
A gas entrainment (GE) evaluation method presented at the previous symposium can predict a gas core length by applying local instant values (obtained from CFD results) to the extension vortex theory. However, in the GE evaluation method, a surface tension effect was not introduced. Therefore, it is valid to consider that gas core lengths were overestimated. In this study, the prediction accuracy of gas core lengths is improved by introducing the surface tension effects into the GE evaluation method. For that purpose, the mechanical balance between gravitational, centrifugal and surface tension forces are considered. The improved method was validated by predicting the gas core lengths in basic experiments. As the results, the predicted gas core length values by the improved evaluation method gave better agreements with the experimental results than the original evaluation method.
