In the current time the task of management the accumulated volumes of irradiated graphite obtained by uranium-graphite reactors decommission (over 250,000 tons in the world) is acute in Russia along with other countries. It is known during the reactors operation the carbon and chlorine present in the nuclear graphite become activated. Besides various incidents during reactor exploitation occur resulting in contamination of both block and bushing graphite with fission products and actinides. Conditions of the reactor operating determine both properties of irradiated graphite and mechanisms of its pollution. Therefore, a comparative study of the decontamination effectiveness of different species of irradiated graphite is of considerable interest.
Thus the purpose of this work is to study the radionuclides distribution in the volume of decommissioned reactor graphite bushes and to assess the effectiveness of irradiated graphite decontamination by using liquid reagent treatment methods as well. The authors used the samples of irradiated graphite sampled from bushings after two years of operation in the reactors.
Involved graphite samples belonged to the graphite masonry, some zones of which were undergone the incidents. For deactivation experiments the sampled graphite were milled and averaged. Solutions containing acids, alkali, oxidizing reagent, mixtures of acids with fluoride ion were used. Experiments lasted 30 days.
At first stage solutions of 2.5 M H2O2, 5 M NaOH, 1-2 M H3PO4, HCl, HNO3, mixtures of acids with fluoride ion and etc. were used at a temperature of about 22°C. Afterwards were used more aggressive solutions. The efficiency of the deactivation process of bushing graphite samples is minimal in solutions of hydrogen peroxide and alkali, slightly higher in acid solutions, but generally does not exceed 10%, including for Cl-36 and C-14. Increase in the concentration of hydrochloric and nitric acids among with fluoride ions addition gives rise to increasing of Cs-137, Co-60, and Am-241 deactivation degree by almost an order of magnitude. Cl-36 and C-14 extraction degree increases as well. However for Pu-239, U-238 and Sr-90 this effect was not observed. As the temperature rises the recovery of Pu-239 and U-238 increases, while the process time decreases.
It should be noted that an increase in the decontamination time at elevated temperature practically does not affect on the efficiency of Cl-36 and C-14 extraction process.
In the long term taking into account the peculiarity of irradiated graphite properties, it is proposed to evaluate the possibility of using liquid deactivation to reduce the activity of both non-removable and removable graphite elements.