Groundwater flow modeling focused on the Fukushima Daiichi Nuclear Power Plant site

Saegusa, Hiromitsu; Onoe, Hironori; Kohashi, Akio; Watanabe, Masahisa

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company is facing contaminated water issues. The amount of contaminated water is continuously increasing due to groundwater leakage into the underground part of reactor and turbine buildings. Therefore, it is important to understand the groundwater flow conditions at the site and to predict the impact of countermeasures taken for isolating groundwater from the source of the contamination, i.e. the reactor buildings. Installations, such as of land-side and sea-side impermeable walls have been planned as countermeasures. In this study, groundwater flow modeling has been performed to estimate the response of groundwater flow conditions to the countermeasures. From the modeling, groundwater conditions and changes in response to implementation of the countermeasures could be reasonably estimated. The results indicate that the countermeasures will decrease the volume of inflow into underground part of the buildings. This means that the countermeasures will be effective in reducing the discharge volume of contaminated groundwater to ocean.

Analysis of pressure- and temperature- induced steam generator tube rupture during PWR severe accident initiated from station blackout

Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00

Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.

Improvement of CHF correlations for research reactors using plate-type fuels

Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio

Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1815 - 1822, 1997/00

In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as "Loss of the primary coolant flow". Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and nonuniform heat flux condition. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed.

Study of two-phase flow under low velocity in PWR-LOCA

Onuki, Akira

JAERI-M 92-150, 134 Pages, 1992/10

JAERI-M-92-150.pdf:4.08MB

no abstracts in English

BETHSY/LSTF counterpart test on natural circulation in a pressurized water reactor

P.Bazin*; R.Deruaz*; Yonomoto, Taisuke; Kukita, Yutaka

ANS Proc. of the 1992 National Heat Transfer Conf., p.301 - 308, 1992/00

no abstracts in English

Experimental study of falling water limitation under counter-current flow in the vertical rectangular channel; CCEL correlation for rectangular channel

*; Kaminaga, Masanori; Sudo, Yukio

JAERI-M 88-134, 26 Pages, 1988/07

JAERI-M-88-134.pdf:0.8MB

no abstracts in English

Fundamenntal Experiments on Closed Thermosyphon for JRR-3 Cold Neutoron Source(I)

; *; ;

JAERI-M 87-006, 19 Pages, 1987/02

JAERI-M-87-006.pdf:2.32MB

no abstracts in English

Experimental study of effects of upward steam flow rate on quench propagation by falling water film

Abe, Yutaka; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 23(5), p.415 - 432, 1986/00

https://doi.org/10.3327/jnst.23.415

no abstracts in English

Experimental study of differences in DNB heat flux between upflow and downflow in a vertical channel

Sudo, Yukio; *; ; Kaminaga, Masanori;

Journal of Nuclear Science and Technology, 22(8), p.604 - 618, 1985/00

https://doi.org/10.3327/jnst.22.604

no abstracts in English

Analysis of ROSA-III Test RUN 704 by RELAP5/MOD 0 Code

*; Tasaka, Kanji

JAERI-M 9476, 60 Pages, 1981/05

JAERI-M-9476.pdf:1.78MB

no abstracts in English