Phantom dosimetry and cancer risks estimation undergoing 6 MV photon beam by an Elekta SL-25 linac.

The High-energy linear accelerator (linac) is a valuable tool and the most commonly used devices for external beam radiation treatments in subjects suffer from cancer. To estimate the dose deposited in several organs of a female patient due to pelvic irradiation by an Elekta SL-25 linac in 6 MV photon beam mode, the MCNPX code is used considering the most details of linac. The equivalent dose in different organs is computed according to the face down position (prone) of MIRD and UFRO phantoms. The data obtained using MCNPX show that the received dose in all commons organs of MIRD and UFRO phantoms is 535.73 and 433.09 mSv/Gyx, respectively. The risks of second cancer incidence and mortality during radiotherapy treatment are compared between MIRD and UFRO phantoms. The results indicated that bladder has the maximum risk of secondary cancer incidence risk of 142.85 and 135.34 per 105 persons based on MIRD and UFRO phantoms, respectively; while the total risk is about 1 in 163 and about 1 in 101 in these phantoms.

Specifying the flux and dose-equivalent buildup factors for infinite slabs irradiated by radionuclide neutron sources.

In this work, flux and neutron dose-equivalent buildup parameters are calculated for six radionuclide point-like neutron sources having broad energy spectrum which irradiate infinite slab-like common shielding materials of beryllium, concrete, iron, graphite, water, and lead, employing the MCNPX code. The description of the buildup factor is made in a straightforward way which is analogous to that of gamma. The parameters are obtained for thicknesses of shield from 0.5 to 10 mean free paths (mfp). The achieved dose-equivalent buildup parameters are parameterized by polynomial expressions. Using this parameterization, one can determine these factors for the desired thickness of shield material and each neutron source.

Effect of each component of a LINAC therapy head on neutron and photon spectra.

Linear accelerators (LINACs) are widely applied in radiotherapy for their versatility and flexibility. Monte Carlo simulations were made to find the neutron and photon spectra at the isocenter (IC) of a LINAC operating at 10, 15, 18, and 24 MV by the MCNPX code. A detailed model of the LINAC head, consisting of flattening filter, secondary collimator, primary collimator, and multi-leaf collimator were used in the calculations. The effect of eliminating any of these components on contamination of a neutron spectrum and a photon spectrum was assessed. Photon and neutron ambient equivalent doses were found, and comparisons were made for the various structures. Lethargy neutron spectra at the IC were compared with spectra computed with the function reported by Tosi et al., which describes well neutron spectra for the energy region beyond 1 MeV, although tending to undervalue energy spectra below 1 MeV. The findings show that the photon and neutron fluences are enhanced when eliminating a LINAC component. The neutron and photon doses increased except when removing the primary collimator.

Estimation of photoneutron dosimetric characteristics in tissues/organs using an improved simple model of linac head.

Neutrons as a one of the by-products of high-energy photons in radiotherapy increase the patient's risks and could cause secondary cancers. Therefore, a new corrected simplified model for linac head was introduced to calculate equivalent (H) and absorbed (D) doses in different tissues/organs of a phantom model. The photoneutron spectrum calculated with this model was in agreement with the spectrum obtained by using a detailed linac´s head model having all components. Besides, an anthropomorphic phantom was irradiated under different gantry angles of the corrected simplified model aiming to emulate 15, 18, 20, and 25MV Siemens linacs. The results indicated that tissues which were within the treatment field received more dose than others. Furthermore, tissues which were in the vicinity of each other and the same depth in the phantom nearly received the same doses. Finally, fatal secondary cancer risk was also studied.

Analysis of neutron scattering components inside a room with concrete walls.

This paper describes the scattering corrections needed when neutron detectors are calibrated with a neutron point source at the center of a calibration room. The independence of scattering value from the geometric shape of the room was studied, which for more confidence in this case, two sets of rooms with same inner surface area were evaluated. The parameters that relate the air scattering part (A) and room-return part (R) with the additional contribution from scattered neutrons have been calculated for neutrons whose energy goes from 10-8 to 20MeV. These parameters were calculated using Monte Carlo method for 150, 200, 300, 407, 500, 800 and 1000cm-radius spherical cavity containing air. In the calculations, monoenergetic neutron sources were placed at the center of cavity, and then neutron fluences were determined at several distances of source to detector along the spherical cavity radius. The parameter R has been fitted as a function of surface area (or radius) of spherical room, and the related coefficient has been calculated for 16 monoenergetic neutron sources. It may produce a reasonable estimate (with difference <10%) for the contribution of walls scattering in any geometry of a calibration room.

Improvement in the practical implementation of neutron source strength calibration using prompt gamma rays.

In this study, the neutron emission rate from neutron sources using prompt gamma rays in hydrogen was determined, and several improvements were applied. Using Monte Carlo calculations, the best positions for the source, moderator and detector relative to each other were selected. For (241)Am-Be and (252)Cf sources, the sizes for polyethylene spheres with the highest efficiency were 12- and 10-inch, respectively. In addition, a new shielding cone was designed to account for scattered neutrons and gamma rays. The newly designed shielding cone, which is 45 cm in length, provided suitable attenuation for the source radiation.

Improvement in the calibration time of absolute emission rate of radioisotopic neutron sources using vanadyl sulfate bath.

In this study, for calibration of emission rate of radioisotopic neutron sources, a suitable vanadium salt was proposed instead manganese sulfate because the shorter half-life of (52)V would facilitate faster neutron yield measurements to be made with a shorter delay time between subsequent measurements. Using Monte Carlo method, different correction factors of manganese and vanadyl sulfate baths were calculated and compared with each other. The results showed that for having an appropriate efficiency of VBS, high concentrations of solution must be used.

An evaluation of a manganese bath system having a new geometry through MCNP modelling.

In this study, an approximate symmetric cylindrical manganese bath system with equal diameter and height was appraised using a Monte Carlo simulation. For nine sizes of the tank filled with MnSO(4).H(2)O solution of three different concentrations, the necessary correction factors involved in the absolute measurement of neutron emission rate were determined by a detailed modelling of the MCNP4C code with the ENDF/B-VII.0 neutron cross section data library. The results obtained were also used to determine the optimum dimensions of the bath for each concentration of solution in the calibration of (241)Am-Be and (252)Cf sources. Also, the amount of gamma radiation produced as a result of (n,γ) the reaction with the nuclei of the manganese sulphate solution that escaped from the boundary of each tank was evaluated. This gamma can be important for the background in NaI(Tl) detectors and issues concerned with radiation protection.