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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.

PurposeThis study aims to characterize the neutron radiation field inside a scanning proton therapy treatment room including the impact of different pediatric patient sizes.Materials and MethodsWorking Group 9 of the European Radiation Dosimetry Group (EURADOS) has performed a comprehensive measurement campaign to measure neutron ambient dose equivalent, H*(10), at eight different positions around 1-, 5-, and 10-year-old pediatric anthropomorphic phantoms irradiated with a simulated brain tumor treatment. Several active detector systems were used.ResultsThe neutron dose mapping within the gantry room showed that H*(10) values significantly decreased with distance and angular deviation with respect to the beam axis. A maximum value of about 19.5 µSv/Gy was measured along the beam axis at 1 m from the isocenter for a 10-year-old pediatric phantom at 270° gantry angle. A minimum value of 0.1 µSv/Gy was measured at a distance of 2.25 m perpendicular to the beam axis for a 1-year-old pediatric phantom at 140° gantry angle.The H*(10) dependence on the size of the pediatric patient was observed. At 270° gantry position, the measured neutron H*(10) values for the 10-year-old pediatric phantom were up to 20% higher than those measured for the 5-year-old and up to 410% higher than for the 1-year-old phantom, respectively.ConclusionsUsing active neutron detectors, secondary neutron mapping was performed to characterize the neutron field generated during proton therapy of pediatric patients. It is shown that the neutron ambient dose equivalent H*(10) significantly decreases with distance and angle with respect to the beam axis. It is reported that the total neutron exposure of a person staying at a position perpendicular to the beam axis at a distance greater than 2 m from the isocenter remains well below the dose limit of 1 mSv per year for the general public (recommended by the International Commission on Radiological Protection) during the entire treatment course with a target dose of up to 60 Gy. This comprehensive analysis is key for general neutron shielding issues, for example, the safe operation of anesthetic equipment. However, it also enables the evaluation of whether it is safe for parents to remain near their children during treatment to bring them comfort. Currently, radiation protection protocols prohibit the occupancy of the treatment room during beam delivery.

The MARIA research reactor is designed and operated as a multipurpose nuclear installation, combining material testing, neutron beam experiments, and medical and industrial radionuclide production, including molybdenum-99 (99Mo). Recently, after fuel conversion to LEU and rejuvenation of the staff while maintaining their experience, MARIA has been used to respond to the increased interest of the scientific community in advanced nuclear power studies, both fission and fusion. In this work, we would like to introduce MARIA’ s capabilities in the irradiation technology field and how it can serve future nuclear research worldwide.

This paper presents the results on radiation shielding of special types of concrete, designed for nuclear reactor containments. The research focused on the study of shielding properties against neutron component of ionizing radiation. For this purpose, two types of ionizing source were used: the isotopic reference neutron field of 239Pu-Be in calibration laboratory and the MARIA Research Reactor.