PROCESS FOR THE TREATMENT OF RADIOACTIVE GRAPHITE

Abstract: A process for the treatment of radioactive graphite, which process comprises the steps of i) reacting the radioactive graphite at a temperature in the range of from 250 DEG to 900 DEG C with superheated steam or gases containing water vapour to form hydrogen and carbon monoxide; and ii) reacting the hydrogen and carbon monoxide from step (i) to form water and carbon dioxide. The process enables radioactive graphite, such as graphite moderators, to be treated either in-situ or externally of a decomissioned nuclear reactor.

Description: PROCESS FOR THE TREATMENT OF RADIOACTIVE GRAPHITE The present invention relates to a process for the treatment of radioactive graphite and, in particular, the radioactive graphite which results from the graphite which is used as a moderator material in a number of nuclear reactor designs. Graphite, which consists predominantly of the element carbon, is used as a moderator in a number of nuclear reactor designs, such as the MAGNOX and AGR gas cooled reactors in the United Kingdom and the RBMK design in Russia. During construction the moderator of the reactor is usually installed as an interlocking structure of graphite bricks. At the end of reactor life the graphite moderator, typically weighing about 2,000 tonnes, is a form of radioactive waste which requires eventual management. Graphite is a relatively stable chemical form of carbon, which is in many ways suitable for direct disposal without processing. However, after neutron irradiation the graphite will contain stored igner energy. The potential for release of this energy needs to be accommodated in any strategy which relies on disposing of the graphite in unprocessed form. Alternatively, processing the graphite before disposal can allow the safe release of any stored Wigner energy. The graphite also contains significant quantities of radionuclides from neutron induced reactions, both in the graphite itself and in the minor impurities which it contains. The radioisotope content can conveniently be divided into two categories. Short lived isotopes (such as cobalt-60) make the graphite difficult to handle immediately after reactor shutdown, but decay after a few tens of years. Long- lived isotopes (principally carbon-14) are of concern through the possibility of their discharge to the biosphere. Processing the graphite offers the opportunity to separate the majority of the graphite mass (carbon) from the short-lived radioisotopes . This in turn facilitates management of the graphite waste shortly after the end of reactor life. Because of the characteristics of graphite and its mass, the most common procedure to date for decommissioning of graphite moderated reactors is to "SAFESTORE" the reactor core in-situ for a period of tens of years following reactor shut-down. During this period short-lived radioisotopes decay sufficiently to allow eventual manual dismantling of the graphite moderator. Most plans in the United Kingdom then assume that the graphite will be disposed of in its existing chemical form, with appropriate additional packaging to prevent degradation or release over the long period of carbon-14 decay. Safe-store has certain negative consequences, since it implies a long-term financial liability, a visually intrusive structure that has no productive purpose, a requirement for continual surveillance and the requirement for a future generation (which gained no benefit from the original asset) to complete its eventual clearance. If the safe-store alternative is to be replaced by shorter term management, it is essential for the graphite to be processed in a safe and radiologically acceptable manner. We have now developed a process which provides a convenient means for converting the graphite into a form where any Wigner energy is removed and which can be used to gasify the carbon in the graphite and to practically retain the gasified carbon in a form for further processing as necessary. The process which we have developed can also be used to separate efficiently the carbon in the graphite from other radioactive elements present in the moderator to allow ease of handling, and to facilitate the discharge of the carbon mass to atmosphere. The process can be used to process individual blocks or particles of graphite removed from the reactor core in a conventional manner. The process which we have developed can also be used to react the moderator graphite slowly, in-situ in a slow and controlled manner without the need for human intervention in the reactor core. Accordingly, the present invention provides a process for the treatment of radioactive graphite, which process comprises the steps of i) reacting the radioactive graphite at a temperature in the range of from 250° to 900°C with superheated steam or gases containing water vapour to form hydrogen and carbon monoxide; and ii) reacting the hydrogen and carbon monoxide from step (i) to form water and carbon dioxide. Figure 1 of the accompanying drawings illustrated a flow diagram of a process of treating radioactive graphite according to the invention. The present invention is a process applied to graphite materials previously used as the moderator in the core of a thermal nuclear reactor and which are no longer required for this purpose. It also applies to any other graphite materials (fuel element sleeves, braces etc) irradiated in the neutron flux of a nuclear reactor core. In step (i) of the process of the present invention the reaction of superheated steam or gases containing water vapour with graphite is carried out at a temperature in the range of from 250 to 900°C, preferably 600° to 700°C, to form hydrogen and carbon monoxide. This type of process is generally referred to in the art as "steam reforming". The reaction in step (i) may be carried out with the addition of oxygen to the steam or gases containing water vapour to provide exothermic reaction energy for the process. The addition of oxygen enables the temperature of the steam reforming reaction to be controlled. The gases from step (i) are then further oxidized in step (ii) with oxygen to form carbon dioxide and water. During the process the gases are retained in an enclosed atmosphere. The carbon dioxide and water which are produced in the process may be disposed of subsequently by one of the following procedures: Controlled discharge of the carbon dioxide to the atmosphere, after any further processing necessary to minimise its radioactive content; Compression and liquefaction of the carbon dioxide for temporary storage, processing, transportation or disposal; Condensation of steam to provide water for treatment, disposal or release; and/or Further chemical processing of the carbon dioxide to create a suitable solid waste form for disposal, such as calcium carbonate. The process of the present invention may be carried out either in-situ within the core of a decomissioned nuclear reactor or may be applied (in externally provided equipment) to pieces or particles of graphite which have been removed from the reactor core . Radioactive secondary waste from the process of the invention or from further processing of the carbon dioxide produced in the process prior to discharge, can be dealt with in any conventional manner appropriate to normal procedures of the nuclear plant concerned. In-situ processing of graphite requires that the graphite in the reactor core be subjected to conditions suitable for the gasification of graphite. The in-situ reactions can be performed by various methods, as discussed below. In a first method, carbon dioxide, nitrogen or other inert gas may be recirculated through the reactor using normal in-plant equipment with the addition of small, controlled amounts of steam and/or oxygen, when required. A side stream is continuously extracted from the loop for the removal of carbon monoxide, hydrogen, and carbon dioxide. To mitigate potential hydrogen explosion a catalytic hydrogen converter is preferably inserted in the treatment loop to convert any hydrogen to water. This option requires the injection of a small amount of oxygen into the catalytic converter. For the addition of trace amounts of oxygen, the reactor circuit is maintained above 250°C for oxidation reactions to proceed in sufficient time to allow the use of less than 5 percent oxygen concentration in the recirculating gases. The use of restricted oxygen levels is recommended to eliminate any potential explosion reactions. For the addition of steam, the reactor circuit is maintained above 350°C for the reformation reactions to proceed at reasonable rates. In a second method gases are injected into and removed from the reactor without the use of other in- plant equipment. This method involves the isolation of the reactor from the balance of plant systems. An external gas recirculation loop can be utilized to inject gases into the reactor and provide removal of gaseous reaction products. Selected areas of the reactor can be maintained at high temperature by means of the injection of superheated gases at 400 to 900°C, or by generation of the needed heat inside the reactor. Heat generation inside the reactor can be achieved by the insertion of electrical or combustion tube heaters placed in one or more of the fuel channels. The in-situ reaction utilizing this method allows the preferential removal of the graphite in selected areas of the reactor in order to remove graphite in a planned sequence. This is an extremely valuable safety feature, since it allows the graphite to be removed in a structurally secure manner, avoiding the possibility of collapse of a weakened moderator structure during the later stages of removal. The feasibility of local removal of graphite by this method is further aided by the decreased thermal conductivity of graphite in end-of-life moderators, which occurs as a result of neutron irradiation. It is estimated that over 75 percent of the graphite could be removed this way. The injected gases may consist of an inert gas and steam, together with oxygen, as required. Final removal of the last traces of graphite could, for example, be achieved by reverting to the previous method discussed above. When the process of the present invention is carried out on pieces or particles of the graphite which have been removed from the reactor core, the process may be carried out as a continuous, semi- continuous or batch process. The process may be carried out using a stationary bed formed from the graphite particles or pieces or, preferably, the process is carried out in a fluidized bed reactor. Preferably, the bed will be fluidized using the steam reactant as the fluidization aid, but it will be appreciated that the reactant bed may be fluidized using an inert gas, such as nitrogen or carbon dioxide, with the appropriate injection of steam and/or oxygen, to enable the reaction to proceed. Inert bed material can be utilized in the fluid bed to stabilise temperature where steam and/or oxygen is injected into the vessel. The steam reforming reaction proceeds according to the equation: C + H 2 0 -* CO + H 2 In the second stage of the process of the present invention the carbon monoxide and the hydrogen are oxidised to carbon dioxide and water. This is generally carried out using oxygen gas as the oxidising agent. The oxidation reaction may be carried out in the same vessel as, or a different vessel from, that in which the steam reforming reaction is carried out. For example, when the steam reforming reaction is carried out in a fluidized bed reactor, the oxygen may be introduced into the upper portion of the fluidized bed reactor, so that both steps of the process are carried out in a single reactor vessel. The advantage of the process of the present invention, as compared to the combustion of radioactive graphite, is that it can be carried out under appropriately controlled containment conditions. The loss of hazardous or radioactive materials in the off-gas is therefore reduced or even eliminated. Another significant benefit is the low volume of off- gas that simplifies handling including the possibility of achieving substantially zero gaseous emissions. Further, the process enables the Wigner energy stored in the radioactive graphite to be released in a controlled manner. The present invention will be further described with reference to Figure 1 of the accompanying drawings which is an overview flow diagram of one means of carrying out the process of the present invention. Referring to the drawing, radioactive graphite is remotely removed from a nuclear reactor core by means of water jet or mechanical cutters. Graphite pieces and water are introduced into a size reduction wet grinder 1 where the graphite is reduced to <1.0 cm size. The size reduced graphite is mixed with water in vessel 2 and the slurry is fed directly into the fluidized bed reformer 4 by means of a slurry injector pumps 3, without any other pre-treatment or handling required. Alternatively, graphite can be size reduced to <12.0cm, preferably less than 4.0 cm for direct injection into the reformer 4 by means of a mechanical screw conveyor 5. The fluidized bed reformer 4 serves to evaporate all water from the graphite slurry and other liquid waste feeds and to pyrolyze any organic components through destructive distillation (pyrolysis) . Energy needed to evaporate the feed water and drive the endothermic reformation process is provided by operating the fluid bed in an autothermal steam reforming mode. Carbon monoxide and hydrogen produced by the in-bed steam reformation process are fully oxidised to carbon dioxide and water in the upper zone of the fluid bed by injection of oxygen along line 6. The off-gas from the reformer 4, which leaves the reformer along line 7, contains fine particulates, including most radionuclides and non volatile inorganic materials, such as silica and calcium and gaseous components such as steam, carbon dioxide and gaseous radionuclides, particularly tritium, carbon-14 and iodine. The solid residue is elutriated from the reformer 4 by the fluidising steam and gases. The particulates in the off-gas from the reformer are removed from the off-gas stream by a high temperature filter or wet scrubber 8. If only graphite is to be processed by the process, a high temperature particulate filter is all that is needed to clean all non-volatile radionuclides from the off- gas. If other streams are being processed, the wet scrubber is utilised as shown in Figure 1. Table 1 provides a list of typical radionuclides found in moderator graphite and how the radionuclides are partitioned in the process of the present invention. Table 1 Typical Radionuclide Partition o I Utilisation of the wet scrubber 8 cleans the off- gas by removing particles elutriated from the reformer 4 and neutralises any potential acid gases . The scrubber solution is concentrated by the hot off-gas from the reformer 4 to 1% to 20% by weight solids. The pH in the scrubber solution is controlled between 5.0 and 7.0 to minimise carbon dioxide absorption and to ensure removal of acid gases. The salt solution can be directed along line 9 for treatment by conventional means, such as direct discharge (if radioactivity levels permit) , discharge after selective removal of radioactive species, or encapsulation to form solid waste. Insoluble constituents in the scrubber solution can be removed by filtration if a discharge route is chosen. The warm, water-saturated off-gas stream leaves the scrubber 8 along line 10 and can be further processed to remove essentially all the water vapour by means of a refrigerated condenser 11. The condensed water leaving the condenser 11 along line 12 will include essentially all the tritium from the graphite. The condenser water, with trace levels of tritium, can be handled by one or more of the following methods. It can be recycled to provide for water cutting duty, or to supply superheated steam to the reformer. Alternatively it can be discharged as water vapour or liquid water, or used to mix with cement for solidification of other radioactive waste. Some of the iodine in the off-gas will also tend to be carried with the water leaving the condenser 11 along line 13. The cool, dry off-gas consists almost exclusively of carbon-dioxide and small amounts of oxygen and nitrogen. If allowed by regulation, the carbon- dioxide rich off-gas can be HEPA filtered in vessel 14, monitored at 15, and then discharged to the facility stack at 16. If required by regulation the carbon dioxide can be removed from the off-gas by a refrigerated C0 2 condenser 18. Water absorber 17 is needed to remove any traces of water that could affect operations of C0 2 separator 18. The concentrated carbon dioxide can be transferred along line 20 for conversion into a solid carbonate salt. The remaining non-condensable gases can then be circulated along line 19 to the HEPA filter 14, monitored and then discharged to the facility stack 16. The final small off-gas flow represents less than 5% of the off-gas flow from the outlet of the reformer. The concentrated carbon dioxide stream 20 issuing from the condenser can be converted to a solid, inert carbonate compound using existing technology. For example, the carbon dioxide may be reacted with calcium or magnesium oxides or metals to produce insoluble magnesium or calcium carbonate salts. The interaction of carbon dioxide with MAGNOX fuel element debris waste is described in several publications (for example "CEGB dissolves Magnox fuel element debris at Dungeness" by FH Passant, CP Haigh and ASD Willis Nuclear Engineering International, February 1988 pp 48-51) The carbonate salt can be formed into dense pellets or powder and can conveniently be used to fill void spaces in existing radioactive waste disposal containers .

Claims: CLAIMS : 1. A process for the treatment of radioactive graphite, which process comprises the steps of i) reacting the radioactive graphite at a temperature in the range of from 250° to 900°C with superheated steam or gases containing water vapour to form hydrogen and carbon monoxide; and ii) reacting the hydrogen and carbon monoxide from step (i) to form water and carbon dioxide. 2. A process as claimed in claim 1 wherein the temperature in step (i) is in the range of from 600° to 700°C. 3. A process as claimed in claim 1 or claim 2 where step (i) is carried out in-situ within the core of a decomissioned nuclear reactor. 4. A process as claimed in any one of the preceding claims which is carried out externally of a decomissioned nuclear reactor on pieces or particles of graphite which have been removed from the reactor core . 5. A process as claimed in claim 4 wherein the pieces or particles of graphite which have been removed from the reactor core form the bed of a fixed or fluidized bed reactor. 6. A process as claimed in claim 5 wherein the pieces or particles of graphite have a length of less than 12 cm, preferably less than 4 cm, in their longest dimension. 7. A process as claimed in claim 5 wherein the reactor is a fluidized bed reactor in which the bed is fluidized using steam or a steam/inert gas mixture. 8. A process as claimed in claim 5 wherein steps (i) and (ii) of the process are carried out in a single reactor. 9. A process as claimed in claim 7 wherein the fludized bed contains inert solids. 10. A process as claimed in any one of the preceding claims wherein the oxygen is added to the steam or gases containing water vapour to provide exothermic reaction energy for the process. 11. A process as claimed in any one of the preceding claims wherein the carbon dioxide produced in step (ii) of the process is concentrated and converted to a solid carbonate. 12. A process as claimed in any one of the preceding claims wherein the carbon dioxide is converted to a solid carbonate selected from calcium carbonate and magnesium carbonate.

Inventors: BRADBURY DAVID, MASON J BRADLEY, BRADBURY, DAVID, MASON, J., BRADLEY
Applicants: BRADBURY DAVID, MASON J BRADLEY, BRADBURY, DAVID, MASON, J., BRADLEY
Application year: 2000
Main classification: G21F9
IPC codes: C01B31, C01F5, C01F11, G21F9