Tritium is essential for the evaluation of the exposed dose in the maintenance and repair work in fast breeder reactor (FBR) plants because it transfers more widely than in light water reactor (LWR) plants mainly due to the material difference of the fuel cladding and the coolant.
The tritium transport and trap analysis code, TTT, has been developed by Japan Atomic Energy Agency (JAEA) to estimate tritium and hydrogen behavior in FBR plants. The TTT code was created by Iizawa et al. based on the tritium and hydrogen transport model devised by Kumar. It has been improved by data from the experimental FBR Joyo and the prototype FBR Monju. In the TTT code, it is important to deepen the understanding of hydrogen behavior which affects much for tritium behavior with the mechanisms, such as isotope exchange and coprecipitation between tritium and hydrogen.
In this study, the temperature and time dependence of the hydrogen flux from the steam generator heat transfer tube of Monju in the power rising test performed in 1995 were evaluated using the TTT code. The hydrogen flux was calculated so as to simultaneously fit the measured hydrogen concentration in the primary and secondary coolant and cover gas. The flux which was attributed to the steam oxidation on the water side of the heat transfer tube in the steam generator was found to fluctuate between about 0.1 and 3.1×10−11 g H/cm2/s for the steam generator outlet temperature range from 200 to 480 °C during the reactor operation. The flux appeared to follow the Arrhenius relationship, and it almost agreed with the flux obtained in other plants in the temperature range from 330 to 380 °C. The hydrogen flux in the evaporator decreased gradually with the time of elapse to 30–40 days obeying the parabolic law, and it remained almost constant after the duration. The tendency agreed with the previous results in the 50MW steam generator test facility, which is the large scale sodium facility constructed to perform research and development for Monju.