Measurement of void fraction distribution at high pressure in 4 4 simulated fuel bundle for validation of thermal-hydraulics simulation codes

Ono, Ayako; Nagatake, Taku; Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics and Severe Accidents (SWINTH-2024) (USB Flash Drive), 7 Pages, 2024/06

Japan Atomic Energy Agency (JAEA) is developing a neutronics/thermal-hydraulics coupling simulation code for light-water reactors. Thermal-hydraulic simulation codes applied to the coupling code are expected to calculate the void fraction distribution in a rod bundle under operational conditions, which are necessary for neutron transport simulation, and need to be validated using void fraction distribution data in a rod bundle under high-temperature and high-pressure conditions. Therefore, we have conducted the measurement of the instantaneous void distribution in the 4 4 simulated fuel bundle using a developed wire mesh sensor, which is installed in the pressurized two-phase flow experimental loop of JAEA to obtain the data for code validation.

Development of numerical simulation method of natural convection around heated porous medium by using JUPITER

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

For contaminated water management in decommissioning Fukushima Daiichi Nuclear Power Stations, reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the PCV, it is necessary to examine and evaluate optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work in advance. We have developed a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. Since a large-scale thermal-hydraulics analysis of natural convection is necessary for the method, JUPITER developed independently by JAEA is used. It is however difficult to perform the large-scale thermal-hydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. Therefore, it is possible to analyze the heat transfer of the porous medium by adding porous models to JUPITER. In this study, we report the validation of JUPITER applied the porous model and discuss which heat transfer models are most effective in porous models such as series, parallel and geometric mean models. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the numerical analysis with each model, the numerical result with the geometric mean model was the closest of the models to the experimental results. However, the numerical results of the temperature and the velocity were overestimated for those experimental results. In particular, the temperature near the interface between the porous medium and air was more overestimated.

Development of particle collection technique by using bubble breakup phenomenon in venturi tube

Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Konsoryu, 37(1), p.55 - 64, 2023/03

https://doi.org/10.3811/jjmf.2023.005

In decommissioning Fukushima Daiichi Nuclear Power Station, the issue is confinement of radioactive aerosols in the primary containment vessel. Although a High Efficiency Particulate Air (HEPA) filter is used to collect the aerosol particles, pretreatment equipment such as a scrubber may be applied to reduce the load of HEPA filters. In the scrubber, the aerosol particles are removed by moving from gas to liquid through gas-liquid interface. Since the collection efficiency (CE) depends on gas-liquid interfacial area, fine bubbles are necessary to obtain high collection efficiency. JAEA developed a new particle removal technique by using bubble breakup phenomenon in a Venturi tube. To confirm usefulness of the technique, we performed the CE measurements and observed gas-liquid two-phase flow in the Venturi tube. In comparison with a straight pipe type, the Venturi type can have removed particles more 1,000 than it. The CE is almost the same as a HEPA filter. In addition, the Venturi type has the enough CE as the pretreatment equipment for various materials of particles such as Kanto loam, SUS and oil. Besides, we clarified that the CE of the Venturi type depended on the gas and liquid flow rates. The CE increases with the increase of the liquid flow rate but decreases with the increase of the gas flow rate. This is because the CE is affected by the bubble breakup phenomenon in the Venturi tube. In the experiment, we confirmed that cavitation number which is the parameter of the bubble breakup was related to the CE of the Venturi type.

Sea water flow boiling heat transfer involving sea salt deposition; Role of deposited sea salt

Koizumi, Yasuo*; Uesawa, Shinichiro; Ono, Ayako; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Nihon Kikai Gakkai Netsu Kogaku Konfuarensu 2019 Koen Rombunshu (USB Flash Drive), 1 Pages, 2019/10

no abstracts in English

Measurement of void fraction distribution in a 44 fuel bundle under high pressure condition for validation of two-phase CFD code

Nagatake, Taku; Shibata, Mitsuhiko; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

In the Fukushima Daiichi Nuclear Power Plant accident, reactor cores were cooled by natural circulation due to pump trip. To investigate the accident progress of the Fukushima Daiichi Nuclear Power Plant, it is important to understand the thermal hydraulic behavior in reactor cores including fuel bundles. Flow rate inside cores was relatively low in the natural circulation conditions, then, thermal-hydraulic behavior in the fuel bundles was different from that in the normal operating conditions. To evaluate thermal hydraulic behavior under the accidental conditions, we are developing the numerical simulation codes named TPFIT and ACE3D. These codes are based on two-phase computational fluid dynamics and can simulate the two-phase flow inside fuel bundles including low flow rate condition. Before applying these codes to the thermal-hydraulic behavior, the applicability of these codes must be confirmed. Then, in this study, in order to obtain a validation data for TPFIT and ACE3D code, thermal hydraulic experiment was performed by using test section with a simulated fuel bundle with 44 unheated rods. In this simulated fuel bundle, there were wire mesh sensors, and void fraction distribution data inside the simulated fuel bundle under high pressure condition (max. 2.6 MPa) was obtained. The one of the advantage of wire mesh sensor is that a void fraction distribution of cross section at the same time can be measured. In this paper, void fraction distribution of two-phase flow in a simulated fuel bundle under high pressure condition are reported.

Measurement of water-steam flow behavior and pressure distribution in venturi scrubber

Uesawa, Shinichiro; Horiguchi, Naoki; Suzuki, Takayuki*; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10

Observation of dry patch behavior on copper heat transfer surface and measurement of surface temperature distribution in nucleate pool boiling

Uesawa, Shinichiro; Ono, Ayako; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Dai-55-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 8 Pages, 2018/05

no abstracts in English

Saturated pool nucleate boiling on heat transfer surface with deposited sea salts

Uesawa, Shinichiro; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Journal of Nuclear Engineering and Radiation Science, 3(4), p.041002_1 - 041002_13, 2017/10

https://doi.org/10.1115/1.4036987

Development of numerical simulation method to evaluate heat transfer performance of air around fuel debris, 2; Validation of JUPITER for free convection heat transfer

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

Study on forced-convective boiling heat transfer of seawater with sea salt deposition

Uesawa, Shinichiro; Koizumi, Yasuo; Shibata, Mitsuhiko; Nagatake, Taku; Yoshida, Hiroyuki

Konsoryu, 31(2), p.162 - 170, 2017/06

https://doi.org/10.3811/jjmf.31.162

no abstracts in English