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Kawasaki, Kohei; Ono, Takanori; Shibanuma, Kimikazu; Goto, Kenta; Aita, Takahiro; Okamoto, Naritoshi; Shinada, Kenta; Ichige, Hidekazu; Takase, Tatsuya; Osaka, Yuki; et al.
JAEA-Technology 2022-031, 91 Pages, 2023/02
JAEA-Technology-2022-031.pdf:6.57MB
The document for back-end policy opened to the public in 2018 by Japan Atomic Energy Agency (hereafter, JAEA) states the decommissioning of facilities of Nuclear Fuel Cycle Engineering Laboratories and JAEA have started gathering up nuclear fuel material of the facilities into Plutonium Fuel Production Facilities (hereafter, PFPF) in order to put it long-term, stable and safe storage. Because we planned to manufacture scrap assemblies almost same with Monju fuel assembly using unsealed plutonium-uranium mixed-oxide (hereafter, MOX) powder held in PFPF and transfer them to storage facilities as part of this "concentration" task of nuclear fuel material, we obtained permission to change the use of nuclear fuel material in response to the new regulatory Requirements in Japan for that. The amount of plutonium (which is neither sintered pellets nor in a lidded powder-transport container) that could be handled in the pellet-manufacturing process was limited to 50 kg Pu or less in order to decrease the facility risk in this manufacture. Therefore, we developed and installed the "MOX weighing and blending equipment" corresponding with small batch sizes that functioned in a starting process and the equipment would decrease handling amounts of plutonium on its downstream processes. The failure data based on our operation and maintenance experiences of MOX fuel production facilities was reflected in the design of the equipment to further improve reliability and maintainability in this development. The completed equipment started its operation using MOX powder in February 2022 and the design has been validated through this half-a-year operation. This report organizes the knowledge obtained through the development of the equipment, the evaluation of the design based on the half-a-year operation results and the issues in future equipment development.
Ono, Hitomi*; Takenaka, Keisuke*; Kita, Tomoaki*; Taniguchi, Masashi*; Matsumura, Daiju; Nishihata, Yasuo; Hino, Ryutaro; Reinecke, E.-A.*; Takase, Kazuyuki*; Tanaka, Hirohisa*
E-Journal of Advanced Maintenance (Internet), 11(1), p.40 - 45, 2019/05
Uesawa, Shinichiro; Liu, W.; Jiao, L.; Nagatake, Taku; Takase, Kazuyuki; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 15(4), p.183 - 191, 2016/12
no abstracts in English
Jiao, L.; Liu, W.; Nagatake, Taku; Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki; Takase, Kazuyuki*
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 11 Pages, 2016/10
4 rod bundleLiu, W.; Jiao, L.; Nagatake, Taku; Shibata, Mitsuhiko; Komatsu, Masao*; Takase, Kazuyuki*; Yoshida, Hiroyuki
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 10 Pages, 2016/10
To contribute the clarification of the Fukushima Daiichi Accident, Japan Atomic Energy Agency (JAEA) has been performed experiments to obtain void fraction distribution data, including detailed bubble information such as bubble velocity and size, in steam-water two-phase flow in rod bundle geometry under high pressure and high temperature condition, focusing on low flow rate at the core natural circulation flow condition after the reactor scram. In this research, experimental apparatus for measuring void fraction distribution in the 44 rod bundle was constructed. To measure the void fraction distribution under high pressure and high temperature condition (up to 2.8 MPa, 232 
C), two wire mesh sensors (WMSs) were installed. To confirm the applicability of the installed WMSs and the measuring system for two-phase flow in rod bundle, experiments in air-water two-phase flow under atmospheric pressure and room temperature were performed. As a result, it was confirmed that the installed WMSs can be applicable to the two-phase flow in rod bundle. Measured results, such as instantaneous and time-averaged void fraction distribution in the rod bundle, average void fraction across the cross section of the flow channel, bubble length and velocity, were also reported.
Nagatake, Taku; Takase, Kazuyuki*; Yoshida, Hiroyuki
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/06
no abstracts in English
Yamashita, Susumu; Tokushima, Kazuyuki; Kurata, Masaki; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 5 Pages, 2016/06
In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi Nuclear Power Station, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. At the moment, multicomponent analysis method which can be treated any number of component as a fluid or solid body, Zr-water reaction model and simple radiation heat transfer model were implemented and showed that multicomponent melt flow and its solidification were confirmed in the simplified core structure system. However, the validation of the JUPITER using high temperature molten material has not been performed yet. In this paper, in order to evaluate the validity of the JUPITER, especially, for high temperature melt relocation experiment, we compared between numerical and experimental results for that system. As a result, qualitatively reasonable result was obtained. And also we performed melt relocation simulation on actual core structures designed by three dimensional CAD (Computer-Aided Design) and then we estimated phenomena which might be actually occurred in SAs.
4 bundle at 2 MPaLiu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki
Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06
To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 4 bundle under 1.6 MPa (202 C), 2.1 MPa (215 C) and 2.6 MPa (226 C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.
Takase, Kazuyuki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 57(11), P. 747, 2015/11
no abstracts in English
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11
Yoshida, Hiroyuki; Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11
Yamashita, Susumu; Tokushima, Kazuyuki; Kurata, Masaki; Takase, Kazuyuki; Yoshida, Hiroyuki
Nihon Kikai Gakkai Dai-28-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), 3 Pages, 2015/10
no abstracts in English
Kitamura, Tatsuaki*; Sakamoto, Kensaku; Takase, Kazuyuki
Kashika Joho Gakkai-Shi, 35(Suppl.2), p.59 - 60, 2015/09
no abstracts in English
Sakka, Taku*; Jiao, L.; Uesawa, Shinichiro; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09
no abstracts in English
Jiao, L.; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.387 - 388, 2015/06
In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems. In this study, a numerical simulation of bubbly flow in a vertical square channel was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Matos et al. (2004). Comparisons between the experimental and numerical data revealed, in general, good agreement except serious bubble coalescence appeared in numerical simulation.
Koizumi, Yasuo; Takahashi, Kazuki*; Uesawa, Shinichiro; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-52-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu (CD-ROM), P. 2, 2015/06
Pool nucleate boiling heat transfer experiments were performed for water at 0.101 MPa to examine the elementary process of the nucleate boiling. The copper printed circuit board of a 1.57 mm thick bakelite plate coated with a 0.035 mm thick copper membrane was used for a heat transfer surface. The size of the heat transfer surface was 10 mm 10 mm. direct current was supplied to it to heat it up. The bakelite plate of the backside of the copper layer was taken by 7 mm 10 mm. The instantaneous variation of the backside temperature of the heat transfer surface was measured with an infrared radiation camera. The time and the space resolution of the infrared cameras used in experiments were 120 Hz and 0.315 mm 0.315 mm, respectively. Surface temperatures just before the burn-out measured with 120 Hz suggest that the surface temperature was steadily low at a large part of the heat transfer surface. A small hot-dry area came out at the critical heat flux condition. Then, this small hot-dry area iterated to expand and shrink and gradually grew. Other area was still wetted and kept at low temperature. Eventually the small hot-dry area started to grow continuously and a whole part of the heat transfer surface became hot-dry to reach the physical burn-out. The heat transfer surface was divided into two large areas; the hot-dry area and the low-temperature wetted area until the physical burn-out. The local surface heat flux variation derived from measured surface temperature variation clearly illustrated that the boundary between the dried area and the wetted area moved back and forth and the dried arear gradually grew to reach physical bourn-out at the critical heat flux condition.
Yamashita, Susumu; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In accidents of the Fukushima Daiichi Nuclear Power Plants, by stop of the emergency core cooling system, fuel rods were overheated due to the radioactive decay heat and the oxidization of fuel cladding. Although it is inferred that the core degradation occurred, condition inside the core still has not been revealed. Especially, in order to precisely understand the accumulation condition of debris in lower plenum, detailed and phenomenological relocation process of molten fuel is quite important. In this problem, since an experiment is extremely difficult, numerical simulation will be useful tool for investigating conditions in reactor core. However, existing codes can not be phenomenologically treated relocations process. Therefore a phenomenologically-based numerical simulation method for predicting the melting core behavior including solidification and relocation based on the computational fluid dynamics has been developed in JAEA. Last paper, ICONE 23, in order to distinguish a fuel component which possesses a heat source such as a decay heat from structures which does not possess a heat source such as a core plate, control rod guide tubes and so on, we developed three phases and three component multiphase flow simulation code and performed preliminary analysis of molten core relocation behavior using simplified core structures. As a result, we obtained reasonable results. However, we have not carried out validations for the numerical code yet. In this paper, we show that the results of the numerical test for evaluating the validity of the numerical code and also show that the development of the radiation heat transfer model and its preliminary analysis. In addition, we will report the preliminary analysis of the relocation behavior of molten materials in the simplified core support structure.
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Nagatake, Taku; Yoshida, Hiroyuki; Takase, Kazuyuki; Kurata, Masaki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Koizumi, Yasuo; Takahashi, Kazuki*; Uesawa, Shinichiro; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of 9th International Conference on Boiling and Condensation Heat Transfer (Boiling & Condensation 2015) (DVD-ROM), 10 Pages, 2015/04
Pool nucleate boiling heat transfer experiments were performed for water at 0.101 MPa to examine the elementary process of the nucleate boiling. The copper printed circuit board of a 1.57 mm thick Bakelite plate coated with a 0.035 mm thick copper membrane was used for a heat transfer surface. The size of the heat transfer surface was 10 mm 10 mm. Direct current was supplied to it to heat it up. The Bakelite plate of the backside of the copper layer was taken by 7 mm 10 mm. The instantaneous variation of the backside temperature of the heat transfer surface was measured with an infrared radiation camera. The time and the space resolution of the infrared cameras used in experiments were 120 Hz and 0.315 mm 0.315 mm, respectively. Surface temperatures just before the burn-out measured with 120 Hz suggest that the surface temperature was steadily low at a large part of the heat transfer surface. A small hot-dry area came out at the critical heat flux condition. Then, this small hot-dry area iterated to expand and shrink and gradually grew. Other area was still wetted and kept at low temperature. Eventually the small hot-dry area started to grow continuously and a whole part of the heat transfer surface became hot-dry to reach the physical burn-out. The heat transfer surface was divided into two large areas; the hot-dry area and the low-temperature-wetted area until the physical burn-out. The local surface heat flux variation derived from measured surface temperature variation clearly illustrated that the boundary between the dried area and the wetted area moved back and forth and the dried arear gradually grew to reach physical bourn-out at the critical heat flux condition.
