Investigation of reactor graphite processing with the carbon-14 retention.

2002 ◽  
Vol 757 ◽  
Author(s):  
M. I. Ojovan ◽  
O. K. Karlina ◽  
V. L. Klimov ◽  
G. A. Bergman ◽  
G. Yu. Pavlova ◽  
...  

ABSTRACTThe system C – Al – TiO2 has been demonstrated to be a strong candidate for the processing of irradiated reactor graphite waste with the retention of biologic hazardous carbon-14 in chemically and thermal stable corundum-carbide ceramics. The corundum-carbide ceramics is obtained from the powdered precursors blend through self-sustaining thermochemical reactions. Investigations of the system C – Al – TiO2 were carried out both theoretically and experimentally. The refining thermodynamic calculations of the phase composition of resulting end product were performed for a wide variety of components content in the system being investigated. Aluminium oxycarbides production was taken into account in the calculations. Thermodynamic functions of aluminium oxycarbides Al4O4C and Al2OC have been calculated for this purpose using currently available literature evidences and own assessments of missing data. On the basis of thermodynamic simulation the proportions of the source substances were determined, which result in the aluminium oxycarbides production. These simulation results have been supported by XRD-analysis of produced specimens. The experimental processing of reactor graphite was conducted by the use of self-sustaining reactions in C – Al – TiO2 powder blends. Test specimens were produced by mass ranging from 0.1 to 3 kg in the argon atmosphere. Various techniques were used to characterize the produced specimens. The compressive strength of specimens of corundum-carbide matrices produced ranges from 7 to 13 MPa. The leaching rates of Cs-137 and Sr-90 from specimens ranged between 10-4 and 10-5 g/(cm2.day) respectively. The carry-over of the carbon combined in carbon monoxide from the reacting mixtures during exothermic process may run up to 1% wt. that appropriates roughly to less than 0.01% wt. of the carbon-14 in the irradiated reactor graphite.

Author(s):  
Serge A. Dmitriev ◽  
Olga K. Karlina ◽  
Vsevolod L. Klimov ◽  
Micheal I. Ojovan ◽  
Galina Yu. Pavlova ◽  
...  

The system C–Al–TiO2 is of considerable interest for the processing of irradiated reactor graphite waste with the retention of biologic hazardous carbon-14. Investigations of this system were conducted both theoretically and experimentally. Previously, the thermodynamic calculations of the phase composition of resulting end product were performed for a wide variety of components content in the system being investigated. These simulation results have been supported by XRD-analysis of produced specimens. The experimental processing of reactor graphite was conducted by the use of self-sustaining reactions in C–Al–TiO2 mixtures. A search of modifier additives was performed to perfect end product properties. Test specimens were produced by mass ranging from 0.2 to 3 kg in the argon atmosphere. Various techniques were applied to characterize the produced specimens. The compressive strength of specimens of doped carbide-corundum matrices synthesized ranged from 7 to 18 MPa. The carry over of Cs-137 and Sr-90 during synthesis reaction was about 3% wt. The leachability attained of Cs-137 and Sr-90 from specimens was around 10−5 g/(cm2.day). The carbon-14 is combined in the end product in chemically and thermic stable titanium carbide. The carry-over of the carbon combined in carbon monoxide from the reacting mixtures during exothermic process was less than 1% wt. This corresponds roughly to up 0.01% wt. of the carbon-14 inventory, which can be present in the irradiated reactor graphite.


2015 ◽  
Vol 79 (6) ◽  
pp. 1495-1503 ◽  
Author(s):  
Charalampos Doulgeris ◽  
Paul Humphreys ◽  
Simon Rout

AbstractCarbon-14 (C-14) is a key radionuclide in the assessment of a geological disposal facility (GDF) for radioactive waste. In the UK a significant proportion of the national C-14 inventory is associated with reactor-core graphite generated by the decommissioning of the UK's Magnox and AGR reactors.There are a number of uncertainties associated with the fate and transport of C-14 in a post-closure disposal environment that need to be considered when calculating the radiological impacts of C-14-containing wastes. Some of these uncertainties are associated with the distribution of C-14-containing gaseous species such as 14CH4 and 14CO2 between the groundwater and gaseous release pathways. As part of the C14-BIG programme, a modelling framework has been developed to investigate these uncertainties. This framework consists of a biogeochemical near-field evolution model, incorporating a graphite carbon-14 release model, which interfaces with a geosphere/biosphere model. The model highlights the potential impact of the microbial reduction of 14CO2 to 14CH4, through the oxidation of H2, on C-14 transport. The modelling results could be used to inform the possible segregation of reactor graphite from other gasgenerating wastes.


Radiocarbon ◽  
2018 ◽  
Vol 60 (6) ◽  
pp. 1839-1848
Author(s):  
Dalia Grigaliuniene ◽  
Povilas Poskas ◽  
Raimondas Kilda ◽  
Asta Narkuniene

ABSTRACTThere are two units with RBMK-1500 type reactors at the Ignalina Nuclear Power Plant (Ignalina NPP) in Lithuania where graphite was used as a neutron moderator and reflector. These reactors are now being decommissioned, and Lithuania has to find a solution for safe irradiated graphite disposal. It cannot be disposed of in a near surface repository due to large amounts of 14C (radiocarbon, carbon-14); thus, a deep geological repository (DGR) is analyzed as an option. This study had the aim to evaluate 14C migration from the RBMK-1500 irradiated graphite disposed of in a potential DGR in crystalline rocks taking into account the outcomes of the research performed under the collaborative European project CAST (CArbon-14 Source Term) and to identify the potential to reduce the conservatism in the assumptions that was introduced in the lack of data and led in the overestimated 14C migration. The information gathered during the CAST project was used to model 14C transport in the near field by the water pathway and to perform uncertainty analysis. The study demonstrated that more realistic assumptions could reduce the estimated 14C flux from the near field by approximately one order of magnitude in comparison with the previous estimations based on very conservative assumptions.


2005 ◽  
Vol 345 (1) ◽  
pp. 84-85 ◽  
Author(s):  
O.K. Karlina ◽  
V.L. Klimov ◽  
M.I. Ojovan ◽  
G.Yu. Pavlova ◽  
S.A. Dmitriev ◽  
...  

2008 ◽  
Vol 1107 ◽  
Author(s):  
Olga K. Karlina ◽  
Vsevolod L. Klimov ◽  
Galina Yu. Pavlova ◽  
Michael I. Ojovan

AbstractThermochemical processing of reactor graphite waste is based on self-sustaining reaction 4Al + 3TiO2 + 3C = 3TiC + 2Al2O3 which chemically binds 14C from the irradiated graphite in the titanium carbide. Thermochemical processing was investigated to analyse the behaviour of rare earth elements (REE), where REE = Y, La, Ce, Nd, Sm, Eu and Gd. Both thermodynamic simulations and laboratory scale experiments were used. The REEs in the irradiated reactor graphite are formed as activation products of impurities and spread over the graphite bricks surfaces as well as arise from fission of nuclear fuel. REEs can be used also to substitute for waste actinides as well as to increase the durability of carbide-corundum ceramics relative to waste actinides.Thermodynamic calculations and X-ray diffraction analysis of ceramic specimens synthesized revealed that durable REE's aluminates with perovskite, β-alumina and garnet structures are formed by interaction of REE oxides with the Al2O3 melt during the selfpropagating reaction of ceramic formation.The porous carbide-corundum ceramics synthesized have a high hydrolytic durability, e.g. the normalised leaching rates of 137Cs, 90Sr and Nd are of the order of 10–7 – 10–8 g/(cm2·day).


2009 ◽  
Vol 1193 ◽  
Author(s):  
Nikolai M. Barbin ◽  
Dmitri I. Terentiev ◽  
Sergei G. Alekseyev ◽  
Marat A. Tuktarov ◽  
A. A. Romenkov

AbstractGraphite is used as the neutron moderator and reflector in many nuclear reactors. Obsolete graphite nuclear reactors are put out of operation, leading to formation of a large quantity of radioactive graphite waste.It is proposed that irradiated reactor graphite is processed by high-temperature chemical oxidation in salt melts with an oxidant, which is part of the salt melt, leading to formation of exhaust gases: gaseous compounds of carbon and oxygen (CO2 and CO).This study deals with carbon oxidation and physical-chemical transformations of radioactive elements during the interaction between graphite waste of the atomic power industry and salt melts. The method of thermodynamic simulation is used. The carbon melt decreases the transfer of radionuclides to the gaseous phase as compared to incineration of graphite in the atmosphere.


Author(s):  
Michael I. Ojovan ◽  
Olga K. Karlina ◽  
Vsevolod L. Klimov ◽  
Boris G. Trusov ◽  
Galina Yu. Pavlova ◽  
...  

Abstract During operation of uranium-graphite reactors, waste graphite, containing fragments of nuclear fuel and fission products, as well as radioactive zirconium alloy components from fuel assemblies are produced. A large number of experiments should be carried out for the synthesis of appropriate matrix materials for radioactive nuclides that occur in these wastes. For the choice of processing technologies, an approach was used based on the thermodynamic simulation and application of self-sustaining reactions. A preliminary batch compaction and a hot pressing of the end product were not carried out. The end composite matrix product provides strong retention of the carbon-14 and other radionuclides. The processing technologies proposed are rather simple in implementation, can be realized without complex production equipment and energy supply.


2018 ◽  
Vol 3 (1) ◽  
Author(s):  
Iurii Simirskii ◽  
Alexey Stepanov ◽  
Ilia Semin ◽  
Anatoly Volkovich
Keyword(s):  

Author(s):  
K.L. More ◽  
R.A. Lowden ◽  
T.M. Besmann

Silicon nitride possesses an attractive combination of thermo-mechanical properties which makes it a strong candidate material for many structural ceramic applications. Unfortunately, many of the conventional processing techniques used to produce Si3N4, such as hot-pressing, sintering, and hot-isostatic pressing, utilize significant amounts of densification aids (Y2O3, Al2O3, MgO, etc.) which ultimately lowers the utilization temperature to well below that of pure Si3N4 and also decreases the oxidation resistance. Chemical vapor deposition (CVD) is an alternative processing method for producing pure Si3N4. However, deposits made at temperatures less than ~1200°C are usually amorphous and at slightly higher temperatures, the deposition of crystalline material requires extremely low deposition rates (~5 μm/h). Niihara and Hirai deposited crystalline α-Si3N4 at 1400°C at a deposition rate of ~730 μm/h. Hirai and Hayashi successfully lowered the CVD temperature for the growth of crystalline Si3N4 by adding TiCl4 vapor to the SiCl4, NH3, and H2 reactants. This resulted in the growth of α-Si3N4 with small amounts of TiN at temperatures as low as 1250°C.


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