Design and characterization of a novel neutron shield for BNCT in an experimental model of oral cancer in the hamster cheek pouch at RA-3

Our research group at the Radiation Pathology Division of the Department of Radiobiology (National Atomic Energy Commission) has previously demonstrated the therapeutic efficacy of different BNCT protocols to treat oral cancer in an experimental hamster cheek pouch model. In particular, to perform studies in this experimental model at the thermal facility constructed at RA-3, we designed and constructed a shielding device for thermal neutrons, to be able to expose the cheek pouch while minimizing the dose to the rest of the body. This device allowed for the irradiation of one animal at a time. Given the usage rate of the device, the aim of the present study was to design and construct an optimized version of the existing shielding device that would allow for the simultaneous irradiation of 2 animals at the thermal facility of RA-3. Taking into account the characteristics of the neutron source and preliminary biological assays, we designed the shielding device for the body of the animal, i.e. a rectangular shaped box with double acrylic walls. The space between the walls contains a continuous filling of 6Li₂CO₃ (95% enriched in 6Li), approximately 6 mm thick. Two small windows interrupt the shield at one end of the box through which the right pouch of each hamster is everted out onto an external acrylic shelf for exposure to the neutron flux. The characterization of the shielding device showed that the neutron flux was equivalent at both irradiation positions confirming that we were able to design and construct a new shielding device that allows for the irradiation of 2 animals at the same time at the thermal facility of RA-3. This new version of the shielding device will reduce the number of interventions of the reactor operators, reducing occupational exposure to radiation and will make the procedure more efficient for researchers. In addition, we addressed the generation of tritium as a product of the capture reaction in lithium. It was considered as a potential gaseous effluent discharge and the maximum amount that would be generated during normal operation was calculated. The presence of tritium in a sample of irradiated powder was qualitatively confirmed. The amount of tritium generated was within the allowed range even if the total amount produced were discharged as an effluent. Regarding the tritium retained within the shielding box, the maximum energy of the beta particles emitted in tritium decay is only 19 keV. Thus, they are absorbed within the walls of the shielding box and would not affect the operators or the experimental animals. Additionally, after irradiation, the exposure rate in contact with the shielding device was higher than expected. Based on the activation analysis of the irradiated powder, it was possible to infer the presence of a pure positive beta emitter whose semi decay period corresponds to ¹⁸F. The presence of this nuclide could be due to the secondary reaction 16O(t,n)¹⁸F, described for similar conditions, and would explain the activation of 6Li₂CO₃ employed in the shielding device. We are currently analyzing the possibility of varying the composition of the material by chemically transforming Li₂CO₃ to a compound such as LiF. This strategy would avoid intimate contact between oxygen and the tritium generated, and generation of ¹⁸F. Furthermore, LiF would pose an advantage in terms of shielding properties due to its higher density. (author)