Journal of Nuclear Fuel Cycle and Waste Technology

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be 2.10×10-3, 2.97×10-3 and 1.01×10-1 mSv/h, respectively. And those calculated by MCNP-4C are 1.60×10-3, 2.99×10-3 and 7.88×10-2 mSv/h, respectively. The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01mSv/h for the operation area and 0.15mSv/h for the service (maintenance) area.

Keywords

Shielding,Shielding Design,Hot Cell,QAD-CGGP,MCNP,Effective Dose