Shielding calculations for the proposed NISOL Facility at the ORNL Spallation Neutron Source

The design of the proposed NISOL facility has been investigated with respect to shielding. The Monte Carlo code MCNPX was used to obtain the neutron leakage characteristics from a low-density uranium carbide target and copper or gallium beam stops produced by 1 GeV protons. One-dimensional neutron transport calculations have been performed with the ANISN code using the HILO cross section library to verify the bulk shield at several locations in the facility. The dose levels outside the target cell due to radiation streaming through the small straight ion beam pipe leading from the target into the target cell was calculated by applying an additional sequence of MCNPX calculations for the transport of the neutrons that leak through the pipe into the target cell and by applying the discrete ordinates code DORT to transport the neutrons through the concrete walls and ceiling. The calculations confirm that in the present design of the NISOL facility the radiation produced in the targets and the beam stop is in general safely enclosed by the shielding structure. Only a little more material is needed in the direct line of the ion beam to reduce the dose level at the ceiling of the target cell to an acceptable level.