Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz.

To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also speculate on sensitizing alanine to thermal neutrons by adding boron compounds.

Process-oriented dose assessment model for 14C due to releases during normal operation of a nuclear power plant.

Swedish nuclear utility companies are required to assess doses due to releases of radionuclides during normal operation. In 2001, calculation methods used earlier were updated due to new authority regulations. The isotope (14)C is of special interest in dose assessments due to the role of carbon in the metabolism of all life forms. Earlier, factors expressing the ratio between concentration of (14)C in air and in various plants were used. In order to extend the possibility to take local conditions into account, a process-oriented assessment model for uptake of carbon and doses from releases of (14)C to air was developed (POM(14)C). The model uses part of DAISY which has been developed to model the turnover of carbon in crops. [Hansen, S., Jensen, H.E., Nielsen, N.E., Svendsen, H., 1993. Description of the Soil Plant System Model DAISY, Basic Principles and Modelling Approach. Simulation Model for Transformation and Transport of Energy and Matter in the Soil Plant Atmosphere System. Jordbruksförlaget, The Royal Veterinary and Agricultural University, Copenhagen, Denmark]. The main objectives were to test model performance of the former method, and to investigate if taking site specific parameters into account to a greater degree would lead to major differences in the results. Several exposure pathways were considered: direct consumption of locally grown cereals, vegetables, and root vegetables, as well as consumption of milk and meat from cows having eaten fodder cereals and green fodder from the area around the nuclear plant. The total dose of the earlier model was compared with that of POM(14)C. The result of the former was shown to be slightly higher than the latter, but POM(14)C confirmed that the earlier results were of a reasonable magnitude. When full account of local conditions was taken, e.g. as regards solar radiation, temperature, and concentration of (14)C in air at various places in the surroundings of each nuclear plant, a difference in dose between sites of approximately one order of magnitude was found.