Evaluation of radioactivity in the concrete samples from several accelerator facilities

During operation of an accelerator, its buildings and shieldings are activated by secondary particles, such as neutrons, which are produced by nuclear reaction at a target or beam loss points during acceleration and transportation of particles. Although the specific activity of radioisotopes induced in surrounding materials is low, their amounts of activated materials are quite large. Therefore, how to evaluate residual radioactivity is one of important radiation safety issues in releasing the materials to the environment in case of decommissioning and decontamination of accelerator facilities. In order to provide the clearance level of each radioisotope, we have been developed the analytical method of induced radioisotopes, such as gamma-emitters and beta-emitters. We collected concrete samples from seven accelerator facilities. Concrete samples were obtained by boring the concrete wall or floor of several selected points of each facility into 500 mm depths. Each sample was sliced to 20 mm in thickness and pulverized to powder by a stamp mill. Powdered sample of 60 to 110 g was weighed, and put into a plastic bottle (50φ x 59H). The γ-ray measurement of concrete samples was performed by an automatic γ-ray spectrometer. An empty plastic bottle was also measured to subtract background spectrum. Counting time was set to 40.000 sec. The data acquisition and analysis package 'Gamma-studio' provided by SEIKO EG and G was used for the γ-ray measurement. This package consists of a set of complete algorithms for nuclide identification, background subtraction, efficiency corrections and the determination of the radioactivity for each radionuclide. A standard source containing several nuclides was prepared for the efficiency and energy calibrations. The tritium was extracted by heating using an IR furnace and collected in two cold traps cooled by dry ice/ethanol mixture. The sample temperature and heating time were controlled exactly by the programmable controller. Sample about 1g was weighed and put on a quartz boat and inserted into a quartz tube set on the IR furnace. The appropriate condition of tritium extraction was determined by heating temperature at 800 degree C for 30 min under Ar: gas flow of 200 ml/min. Tritium activity was measured with a liquid scintillation counter. We can observed 3H, 60Co, 152Eu and 134Cs in concrete samples of every facility. The activity of 3H was ten times higher than that of 60Co and 152Eu. The maximum radioactivity of them was observed at a depth of 10 cm, and then decrease exponentially to an extremely low level beyond a depth of 50 cm. These isotopes were produced by thermal neutron capture. The activity of 22Na and 54Mn could be measured in the surface region of concrete samples. As these were produced by fast neutron reaction, the activity was decreased exponentially from the surface. In case of electron accelerator, these can be also produce by photonuclear reactions. The correlation between tritium, 60Co and 152Eu activity was investigated by measuring many concrete samples for several accelerator facilities. The results indicate that their activities are strongly correlated each other. So it would also be concluded that the total activity in shielding concrete will be estimated on the basis of the activity of 60Co and 152Eu. Now, we are trying to measure 36Cl, 55Fe and 63Ni, which are very difficult to measure. Our collected database is useful for the legislation of the clearance levels.