Neutronic and thermal estimation of blanket in-pile mockup with Li2TiO3 pebbles

To evaluate exactly temperature distribution in large volume of tritium breeding materials during the blanket in-pile tests with the JMTR, neutronic and thermal calculations were conducted by using Monte Carlo code ‘MCNP’ with nuclear cross section library of ‘FSXLIBJ3R2’ and the transient and steady-state distribution code ‘TRUMP’. From the results of preliminary estimation of temperature distribution in the blanket in-pile mockup, the calculated values were 24–28% higher than the measured values. One of the reasons is due to overestimation of calculated thermal neutron flux.