Volume 22 (2016): Edizione 2 (June 2016)

The main objective of this study was to investigate the Fricke dosimeter water equivalent system for measurement of dosimetric parameters for photon beam. The parameters measured with the Fricke dosimeter were compared to those obtained with an ionization chamber. In this work characteristics for ⁶⁰Co γ-rays of field sizes ranging from 5 × 5 cm² to 20 × 20 cm² are reported. The measurements were carried out in the secondary standard dosimetry laboratory using a collimated ⁶⁰Co gamma source therapy unit. The ⁶⁰Co beam output in terms of absorbed dose to water was obtained as per IAEA TRS 398 recommendations using cylindrical ionization chamber, whose N_D,w has been supplied by the IAEA's reference laboratory. Specific quantities measured include: output factors, peak scatter factor, lateral beam profiles and percentage depth dose. The Fricke dosimeters were irradiated in a water phantom using the suitable poly (methyl methacrylate), PMMA stand. Our results demonstrate that Fricke dosimeter and ionization chamber agree with each other.

For evaluation of the effective focal spot sizes (EFSS), a method suggested by the EN 60336:2005 standard (standard) could be used. In this study we checked whether it is possible to make some deviations from the requirements of the standard without a significant effect on the result. An image receptor with one intensifying screen or two intensifying screens may be used, but the optical value of the slit image shall be in the range of 1.0÷1.4 and the X-ray tube power shall be ranged of about 30%÷50% of the nominal anode input power. A precision scaled magnifier (magnification of 5÷10x and scale of 0.1 mm) may be used for the slit radiogram width measurement instead of a time-consuming scanning of the slit radiogram. These deviations from the requirements of the EN 60336:2005 standard allows to shorten measurement time and to decrease tube current value during X-ray exposures, which reduces the risk of the Xray tube damage.

New capabilities of biomedical accelerators allow for very precise depositing of the radiation dose and imaging verification during the therapy. In addition, computer algorithms calculating dose distributions are taking into account the increasing number of physical effects. Therefore, administration of high dose fractionation, which is consistent with radiobiology used in oncology, becomes safer and safer. Stereotactic radiosurgery (SRS), which is very precise irradiation with high dose fractionation is increasingly widespread use in radiotherapy of prostate cancer. For this purpose different biomedical accelerators are used. The aim of this study is to compare dose distributions for two techniques: VMAT and CyberKnife. Statistical analysis was performed for the two groups of patients treated by VMAT technique (25 patients), and CyberKnife technique (15 patients). The analysis shows that the dose distributions are comparable, both in the treated area (prostate) and in the critical organs (rectum, urinary bladder, femoral heads). The results show that stereotactic radiosurgery of prostate cancer can be carried out on CyberKnife accelerator as well as on the classical accelerator with the use of VMAT technique.

This paper presents the set of procedures developed in Radiation Protection Measurements Laboratory at National Centre for Nuclear Research for evaluation of shielding properties of high performance concrete. The purpose of such procedure is to characterize the material behaviour against gamma and neutron radiation. The range of the densities of the concrete specimens was from 2300 to 3900 kg/m³. The shielding properties against photons were evaluated using ¹³⁷Cs and ⁶⁰Co sources. The neutron radiation measurements have been performed by measuring the transmitted radiation from ²³⁹PuBe source. Scattered neutron radiation has been evaluated using the shadow cone technique. A set up of ionization chambers was used during all experiments. The gamma dose was measured using C-CO₂ ionization chamber. The neutron dose was evaluated with recombination chamber of REM-2 type with appropriate recombination method applied. The method to distinguish gamma and neutron absorbed dose components in mixed radiation fields using twin detector method was presented. Also, recombination microdosimetric method was applied for the obtained results. Procedures to establish consecutive half value layers and tenth value layers (HVL and TVL) for gamma and neutron radiation were presented. Measured HVL and TVL values were linked with concrete density to highlight well known dependence. Also, influence of specific admixtures to concrete on neutron attenuation properties was studied. The results confirmed the feasibility of approach for the radiation shielding investigations.