The mechanics of inserting a photoneutron capability into the standard neutron, photon and electron Monte Carlo transport code, MCNP, is described. Whilst the (gamma,n) cross sections are input as data statements, the angular and energy distributions of the emitted neutrons are taken from simple models. Under the circumstances of bremsstrahlung spectra of end point up to 25 MeV incident on some high Z materials (typical of materials in the accelerator head of medical LINAC's), the energy distribution is well approximated by the Weisskopf nuclear model requiring few parameters and the angular distribution is isotropic. (There exists also a small anisotropic direct component.) Comparisons of photoneutron production with other calculated results and with experimental results showed a relative insensitivity to the parameters employed in the nuclear models. The precise shape of the neutron spectra however showed some sensitivity to the exact value of the threshold of the (gamma,n) cross section. Comparisons were also made between the calculated neutron dose and spectra in the patient plane of medical LINAC's and experimental values. The results were very satisfactory.
|Titolo:||Modelling Photoneutron Reactions in Standard Monte Carlo Codes: Evaluation of Neutron Spectra and Doses from Medical LINAC's|
|Data di pubblicazione:||18-ott-2001|
|Appare nelle tipologie:||4.1 Contributo in Atti di convegno|