Contamination of Graphite Sleeves
Analysis of Radioactive Contamination of the Graphite Sleeves Removed from the Plutonium Production Reactors of SGCE (Tomsk-7)
Tech Area / Field
- ENV-SPC/Solid Waste Pollution and Control/Environment
- ENV-RWT/Radioactive Waste Treatment/Environment
8 Project completed
Senior Project Manager
Tocheny L V
MIFI, Russia, Moscow
- Siberian Chemical Combine, Russia, Tomsk reg., Seversk
- Battelle Memorial Institute, USA, VA, Arlington\nPacific Northwest National Laboratory / Battelle, Putting Technology to Work, USA, WA, Richland\nFRAMATOME, France, Paris La Défense\nUS Department of Energy, USA, DC, Washington
Project summaryThe graphite sleeves are spare details of uranium-graphite reactors of different type. Replacement of the sleeves is carried out each 2-3 years so that for the full time of the reactor operation the mass of irradiated spent sleeves becomes comparable with mass of graphite in the reactor stack. In case of reactors-producers of weapons-grade plutonium (PUGR in Russian abbreviation), the mass of spent graphite sleeves reaches 1000 tons per a reactor.
The most important difference between states of the graphite stack and the graphite sleeves of the reactors decommissioned is caused by difference in conditions of their irradiation and storage. Duration and dates of irradiation mainly define a level and a composition of their radioactive contamination: duration of the sleeves exposure is 10-15 times longer than that of the stack, and the most part of the sleeves was placed in the reactor during the period after the last accident with uranium release from technological channels. An analysis revealed the following another possible essential differences as well:
- the larger relative contribution of superficial contamination to total contamination of the sleeves due to accidental penetration of radionuclides with water-vapour mixture;
- difference in neutron field: fast neutron flux through the sleeves is rather higher while thermal neutron flux is rather less due to flux depression near to the fuel channel. Cross-sections of neutron reactions differed because of temperature difference in the sleeves and in the stack blocks;
- initial graphite compositions were different: the graphite of different sorts and with different composition of impurities was used for fabrication of the sleeves and the blocks;
- the tritium leaching from the sleeves probably occurred under accidents;
- an additional amount of nitrogen might penetrate into the sleeves during the blow-out.
Till now, an experimental information about radioactive contamination of the spent graphite sleeves discharged from the Russian PUGRs is practically absent. Under performance of the ISTC Project No. 561-96, several assays were carried out for the graphite samples taken from the sleeves of the reactors EI-2 and I-1 of the Siberian Chemical Combine (SChC). The measurements were performed with application of b-spectrometry (analysis of 14C, 3H) and g-ray and X-ray spectrometry (analysis of corrosion products, fission products and transuranium elements). The results of assays confirmed difference in contamination between the sleeves and the blocks.
So, the data obtained in assays of the graphite blocks can not be spread on the graphite sleeves.
Because of the differences indicated above, another technology and another duration of treatment must be chosen for utilisation of the graphite sleeves comparing with the graphite blocks.
Spent graphite sleeves were stored at the reactor sites, near to the reactor buildings, in concrete underground or near-to-surface storages of trench type. These storages were constructed in 1950—60-s. Now they do not correspond to current requirements. Therefore, the decommissioning projects of the SChC PUGRs included cleaning of existing storages and re-disposal of radioactive wastes.
Thus, the spent graphite sleeves represent now a huge amount of the radioactive wastes stored under unsatisfactory conditions. Studies of the sleeves contamination, their sorting on the contamination level for utilization (the most part of the sleeves) or for further storage under proper conditions (highly contaminated sleeves) represent an important and urgent problem.
Under performance of the ISTC Project No. 561-96, the technology was developed for taking the graphite samples for assays; MEPhI and SChC laboratories were additionally equipped with measuring and computing techniques; an experience in assaying the reactor graphite was accumulated, and necessary scientific contacts were established. These prerequisites give to the group that works under the Project a capability to solve the following problem: to obtain the necessary information about radioactive contamination of the graphite sleeves stored in the near-to-reactor storages of the SChC and to evaluate the technology of the sleeves utilization by incineration. Rapid control method is needed for sorting the sleeves. The justification task for such a method is included into present project.
The results obtained during performance of the project proposed could be applied to developing the strategy of the graphite management at other decommissioned Russian PUGRs, and, in the future, at the reactors of RBMK type.
The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.
ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.