Reduction of the gamma-ray component from 252Cf fission neutron source--optimization for biological irradiations and comparison with MCNP code

Abstract

Gamma-rays contribute 33% of the absorbed dose from an unfiltered 252Cf fission neutron source. To reduce this gamma-ray component and to enable radiobiological experiments at as high a dose rate as possible, Monte Carlo calculations for several filter materials (Al, Fe, Pb and concrete) have been made using MCNP neutron and photon transport code version 4a. A lead filter of thickness 4 cm was found to reduce the gamma-ray component to 6.7% of the total dose whilst reducing the neutron dose by only about 10%. Such a filter was installed at the MRC 252Cf neutron irradiation facility and dosimetric measurements were made using a TE-TE chamber and a 7LiF(Mg, Cu, P) TLD. Monte Carlo simulations agree with experimental measurements of neutron and gamma-ray doses within 6%. V79-4 Chinese hamster cells were irradiated with lead-filtered and unfiltered neutrons and also with 60Co gamma-rays at two dose rates. The survival fraction obtained for each radiation was consistent with the reduced gamma-ray dose. The relative biological effectiveness for neutrons alone, corrected for gamma-ray effects, was found to be 9.2 +/- 3.4 from the initial slopes and 3.1 +/- 0.5 at 10% survival, both relative to the acute gamma-rays.