Accident scenarios in spent fuel pools (SFP) were not a main topic in nuclear safety studies so far. Heat fluxes are low and the timeframe for counteractions in case of a loss of cooling is long. However, because of their huge activity inventory safety of spent fuel pools should be analyzed with great intensity. Beside some analytical approximations from the U.S. Nuclear Regulatory Commission (NRC) there is only few experimental data about loss of coolant accidents in a SFP [1, 3, 4, 5, 6]. After the reactor accident in Fukushima Daiichi NPP this issue became omnipresent in public perception. The presented paper investigates boil-off scenario of a boiling water reactor (BWR) SFP after a station blackout with focus on experimental findings.

Since 2006 experiments for SFP boil-off scenarios were conducted at the Technische Universität Dresden [7]. For a detailed investigation under realistic geometrical fuel assembly conditions the test facility ADELA-II was built in 2011. It consists of two channels. The inner channel simulates a quarter of a BWR fuel element with a 24 rod bundle. In the outer channel 8 more heating rods were assembled to reduce heat losses from the inner channel and to simulate the surrounding. With a heated length of 3760 mm and a width of 120 mm it was designed in an axial and radial scale of 1:1. The heating profile is based on the heat flux profile of a fuel rod with low burn-up. To measure the detailed axial and radial temperature profiles 113 thermocouples are mounted. With these data conclusions for heat and mass transfer can be made with special regard to convection inside the assembly. Boil-off experiments at the ADELA-II facility with a supplied heat of 20 Watts per rod lead to rod surface peak temperatures of about 479 °C. Due to the high temperature and narrowed geometry the axial heat transport is harmed and radial heat conduction had a great influence on the test results. Accordingly, quasi adiabatic boil-off could not be verified. Thus, coolability of a fuel assembly strongly depends on the (radial) heat removal from the fuel assembly box to the surrounding. Further investigations have to be done for a better consideration of neighbored fuel elements and global thermohydraulic effects in the spent fuel pool.

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