Experimental investigation of spray cooling behavior in 44 simulated fuel bundle

Nagatake, Taku; Shibata, Mitsuhiko; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Journal of Nuclear Science and Technology, 14 Pages, 2022/00

https://doi.org/10.1080/00223131.2022.2096146

In Fukushima Daiichi Nuclear Power Plant accident, failure of cooling system for spent fuel pool occurred and there was a concern that the spent fuels were damaged. Then a safety measures for SFP cooling in severe accident condition is required. As a countermeasure for SFP severe accident, it is considered that a portable spray is used for SFP cooling in such condition. In this research project, the numerical simulation methods have been developed in order to evaluate the applicability of portable spray system for cooling SFPs. And experiments were also performed in order to get a knowledge of spray cooling phenomena and validation data for the numerical simulation methods. As one of the experiments, a cooling experiment by using 44 simulated fuel assembly were performed and temperature distribution during spray cooling process were measured. In this paper, the results of a cooling experiment are reported.

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 7; Analysis on effectiveness of spray cooling by the SAMPSON code

Suzuki, Hiroaki*; Morita, Yoshihiro*; Naito, Masanori*; Nemoto, Yoshiyuki; Nagatake, Taku; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

In this paper, modification of the SAMPSON code was carried out to enable the analysis of spray cooling. The SAMPSON analysis of a spray cooling experiment was performed to confirm reproducibility of spray cooling behavior of fuel claddings. The modified SAMPSON code was applied to a hypothetical loss-of-coolant accident analysis of the SFP. Effectiveness of spray cooling on cladding temperature behavior was investigated. The SAMPSON analysis showed that spraying from the top of the SFP was effective for cooling the fuel assemblies exposed to the gas phase.

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

Study on spray cooling capability for spent fuel pool at coolant loss accident, 1; Research plan

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Koizumi, Yasuo; Yoshida, Hiroyuki; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 4 Pages, 2016/11

The Fukushima Daiichi NPP accident asks that the accident management of the LOCA in the SFPs must be considered to avoid occurrences of severe accident in the SFPs. To prevent the failure of the spent fuel assemblies at the LOCA, transportable spray systems are expected to be put into use to discharge water into fuel assemblies to moderate the temperature increase. To apply the spray system as a countermeasure for the LOCA of the SFP, the capability of the spray cooling system must be evaluated to keep the spent fuel rods safety. JAEA has started the research project to investigate the spray cooling capability for the SFP. In this research project, we aim to construct a numerical simulation method for evaluating the capability of the spray cooling. To develop the method, the basic key phenomena that affect the cooling performance must be clarified and the validation data required for the code development. To clarify the basic key phenomena that affect the cooling performance, that is, the CCFL and the drop size effect on the CCFL, and to obtain the code validation data, we are planning to carry out 2 experiments with two test sections, the spray visualization experiment and the spray cooling experiment. The spray visualization test section aims to get CCFL data in air-water two-phase flow and to understand the two-phase flow behavior over the upper tie plate. The spray cooling test section aims to get the CCFL data in steam-water two-phase flow and to obtain the validation data. This paper focus on the outline of the research plan for the whole research project.

Measurement of void fraction distribution in air-water two-phase flow in a 44 rod bundle

Liu, 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.

Development of numerical method for behavior of fuel melting considering an effect of multi-component

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

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 2; Heat transfer and flow visualization experiment by using internally heated annulus

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

The Thermal-hydraulic behavior of seawater in an internally heated annulus

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

Numerical simulations on melting behavior of simulated fuel elements based on particle method

Nagatake, Taku; Yoshida, Hiroyuki; Takase, Kazuyuki; Kurata, Masaki

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11

Thermal-hydraulic experiments with sodium chloride aqueous solution

Jiao, L.; Liu, W.; Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Nagase, Fumihisa

Proceedings of 15th International Heat Transfer Conference (IHTC 2014) (USB Flash Drive), 11 Pages, 2014/08

In Fukushima Daiichi nuclear disaster, seawater was injected into the nuclear core, which may change the heat transfer characteristics in the reactor pressure vessels (RPV) due to the different physical properties of seawater and pure water. To remove molten fuel from the Fukushima Daiichi Nuclear Power Plants, it is necessary to know the current status of the reactors. Therefore, in this paper, we measured the basic thermal-hydraulic data in an annular tube with a co-axial heater, which includes the heat transfer rate and the pressure drop, using the sodium chloride aqueous solutions and the synthetic seawater as working fluids. The experiments were performed under atmosphere pressure, with the salinity, the fluid mass flux, the inlet temperature and the heat flux used as the parameters. The experimental results and analyses are reported in this paper and the basic influence of the salinity on the heat transfer and the hydraulic characters are proposed.