• 专利附录
  • 专利说明
  • 权利要求
  • 类似技术
  • 同族专利
  • 引用文献
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    • 专利摘要:

    公开号:SE537020C2
    申请号:SE1250386
    申请日:2012-04-18
    公开日:2014-12-02
    发明作者:J Bradley Mason；Sahar Torabzadeh；Thomas Brown；Jonathan Olander
    申请人:Studsvik Inc；
    IPC主号:
    专利说明:

    [4] Due to the properties of graphite and its mass, it is currently the most common method of decommissioning graphite-moderated reactors to store the reactor core in place for about ten years after the reactor is shut down. During this period, short-lived radioisotopes decay sufficiently to allow any manual disassembly of the graphite moderator. Most disposal routes assume that graphite will be disposed of in its existing chemical form, with appropriate additional packaging to prevent degradation or release during the long period of carbon-14 and k | or-36 decomposition |.
    [5] Storage has certain negative consequences, such as the following: 1) a consequence of far-reaching financial responsibility, 2) a visually intrusive storage structure that has no productive purpose and 3) a requirement is placed on future generations (who were not allowed to benefit from the original asset) to complete any elimination. If the storage option were to be replaced by short-term handling, it is necessary for the graphite to be processed in a safe and radiologically acceptable manner.
    [6] Some prior art techniques for treating radioactive graphite added heat and oxidizing gases to treat the graphite to remove a sufficient proportion of the persistent radionuclides within the graphite. These processes have shown that heating or "roasting" alone with inert gases, such as nitrogen or argon, can in itself remove substantially all of the hydrogen-3 (tritium) but the process cannot remove more than sixty (60) percent of carbon-14.
    [7] Test results for these processes show that, if the concentration of oxygen-containing gases is sufficiently limited so that gasification of bulk graphite decreases at temperatures above about 1200 ° C, the removal of carbon-14 substantially decreases to less than about sixty (60) percent, which is unsatisfactory. If the concentration of the oxygen-containing gas is increased so that carbon-14 is satisfactorily removed, too much bulk graphite will be gasified. In any case, the goal of evaporating more than ninety (90) percent of carbon-14, while reducing graphite gasification to less than five (5) weight percent, can not be met by these conventional methods.
    [8] What is needed are systems and methods that can expose graphite to a temperature range sufficient to vaporize the radionuclides without gasifying the bulk graphite, and in particular systems and methods that can remove more than 90 percent of carbon-14 while leaving less than 5 percent. percent of the bulk graphite is gasified.
    [9] Examples of embodiments of the present invention relate to processes that can expose graphite to a temperature range sufficient to vaporize radionuclides without significantly gasifying the bulk graphite. One aspect of the invention relates to a process comprising the steps of (1) heating the grate to a temperature between 1200 ° C to 1500 ° C; (2) introduction of graphite contaminated with radionuclides into the rust; (3) introducing inert gas into the grate; (4) introducing a reducing gas into the grate; and (5) removing evaporating radionuclides from the rust. The process may also comprise the additional steps: o adding oxidizing gas to the rust, and / or reducing the size of the graphite before introducing the graphite into the rust.
    [10] Figure 1 shows a block diagram of a system for treating radioactive graphite according to an embodiment which is an example of the present invention.
    [11] Figure 2 shows a flow chart of the method for treating radioactive graphite according to an embodiment which is an example of the present invention.
    [12] Figure 3 shows a schematic diagram of a rust for treating radioactive graphite according to an embodiment which is an example of the present invention.
    [13] The exemplary embodiments of the present invention provide systems and methods for treating radioactive graphite contaminated with tritium, carbon-14 537 020 and cor-36 and other radionuclides formed during operation of a nuclear reactor or other nuclear process. The systems and processes comprise a grate, which has an operating temperature in the range 800 ° C to 2000 ° C, with inert any oxidizing and reducing gases.
    [14] Figure 1 describes a block diagram of a system 100 for treating radioactive graphite according to an embodiment which is an example of the present invention. Referring to Figure 1, a material handling component 110 receives graphite for processing in the system 100. Typically, the graphite is used as a moderator in a nuclear reactor core. Other graphite sources include, but are not limited to, fuel elements, spacers, or other reactor components irradiated with the neutron flux from the reactor. This graphite usually becomes contaminated with radionuclides such as hydrogen-3 (tritium), carbon-14, cor-36, iron-55, and cobalt-60, and may contain other fission and activation products.
    [15] The material handling component 110 measures and retains the graphite prior to the introduction of the graphite into the grate 120. The graphite received in the material handling component 110 has been removed from the nuclear reactor by any conventional method. Such processes may include wet processes, dry processes, or a combination of both. The present invention can receive either dry or wet graphite of any size or shape resulting from the removal process. Furthermore, the graphite can be soaked in water or other solution before being fed into the material handling component 110.
    [16] The graphite can be treated in granular or powder form. A size-reducing sub-component 112 of the material handling component 110 reduces the size of the received graphite prior to its introduction into the grate 120. In this exemplary embodiment, the received graphite is reduced to a size less than 20 mm. The reduced size increases the volatility of the radionuclides in the graphite. To reduce the size of the graphite, the exemplary subcomponent 112 includes a jaw crusher or rotary crusher.
    [17] The grate 120 comprising a container used to process the sized graphite. The grate 120 has an operating temperature in the range 800 ° C to 2000 ° C. The capacity, shape and size of the grate 120 may vary depending on the application. The grate 120 is constructed of a material suitable for high temperature processes, such as a refractory lined steel container. The operating pressure can vary from high vacuum to slightly pressurized. Any rusting apparatus, including a fluidized bed, movable bed, batch or static rust bed, can be used. An example of a grate is a vertically oriented moving bed grate, where the untreated graphite enters the top of the moving bed and the treated graphite is removed from the bottom of the moving bed while the purge gas flows upwards (countercurrently) through the graphite bed. (See Figure 3, discussed below).
    [18] The grate 120 comprises gas inlets 130, 140, 150 for receiving one or more inert purge gases, one or more reducing gases, and optionally one or more oxidizing gases. Of course, the gas inlets 130, 140, 150 may be a single inlet connected to three different gas sources, a source providing inert purification gas, a second source providing reducing gas, and a third source providing oxidizing gas. The gas inlet or inlets would typically be located near the bottom of the grate 120, so that the gases can enter the container and travel up through the graphite inside the grate. The gas can be introduced through a flow divider or distributor to distribute the gas through the graphite volume, but this component is not a requirement. The grate comprises an outlet 122 for vaporized radionuclides, which are discharged through the outlet 122 of the inert purification gas. The grate 120 also includes an outlet 124 for the treated graphite. 537 020
    [19] Volatile radionuclides are transported out of the grate by the purge gas stream and stabilized in the treatment subsystem 160, using a suitable technique for treating the radionuclides. The treated graphite is further processed in the treatment subsystem 170, whereupon it is packed for final disposal as "clean" (non-radioactive) waste or recycled.
    [20] Carbon-14 is more reactive or more mobile than the co-12 belly of the graphite matrix. The presence of small amounts of oxygen produces the oxygen required to convert carbon-14 to carbon monoxide. The reducing gases suppress the oxidation of carbon-12 in the graphite matrix. An advantage of introducing reducing gas is, for example, that possible carbon compounds in the graphite include cyanide. The introduction of hydrogen into the rust allows hydrogen atoms to bind to cyanide to form hydrogen cyanide, which is volatile, thus removing some carbon-14 through the presence of the reducing gas, including hydrogen.
    [21] Figure 2 shows a flow chart of the method 200 for treating radioactive graphite according to an exemplary embodiment of the present invention. Referring to Figures 1 and 2, at step 210, the graphite is introduced into the grate 120 from the magazine sub-component 114 by the material handling component 110 with a mechanical transfer of the graphite into the grate. In this exemplary embodiment, the process is performed batchwise. Alternatively, the graphite can be treated in a continuous process, such that the graphite enters the top of the grate 120 and exits at the bottom of the grate 120 and the reducing gases enters the bottom of the grate 120 and exits the top of the grate 120. The reservoir component 114 can be omitted.
    [22] Before introducing graphite into the grate 120, the rust is heated to the treatment temperature.
    [23] In step 220, reaction gases are introduced into the grate 120. These gases come into contact with the hot graphite as they flow through the heated graphite. These 537,020 reaction gases comprise at least one inert purification gas and one reducing gas.
    [24] At step 230, the purge gas is collected in the treatment subsystem 160, where the radionuclides are stabilized using known methods. At step 240, the graphite is removed from the grate 120 and treated in the treatment subsystem 170. The treated graphite can typically be deposited in a landfill or recycled and 537,020 would be treated as low-level radioactive waste instead of intermediate-level radioactive waste.
    [25] Figure 3 shows a schematic view of an exemplary rust 300. Graphite is introduced through the feed system (not shown), such as a magazine, at the inlet 310, under a blanket of inert gas. Reaction gases are introduced at the inlet 370 so that the reaction gases flow up through the graphite and out through the gas outlet 320 as the graphite moves down through the container 330. As the graphite moves through the container 330, which may be a ceramic tube, it is heated (shown as heated graphite 340). The container 330 is surrounded by a heat source 350, such as electronic heating coils. The container 330 and the heat source 350 are enclosed in an outer vessel 360, such as a metal shell lined with refractory material. Treated graphite is removed from the container 330 through an outlet opening 380.
    [26] One skilled in the art would appreciate that the present invention provides methods of treating radioactive graphite contaminated with tritium, carbon-14, and chlor-36 and other radionuclides generated during operation of a nuclear reactor or other nuclear process. The processes comprise a grate whose operating temperature is in the range 800 ° C to 2000 ° C with inert gases, optionally oxidizing gases and reducing gases. The combination of temperatures and gases allows the removal of most to substantially all of the carbon 14 in the graphite while substantially gasifying the bulk graphite.
    权利要求:
    Claims (1)
    [1]
    A process comprising the steps of: heating a rust to a temperature between 1200 ° C to 1500 ° C; introduction of graphite contaminated with radionuclides into the rust; introduction of an inert gas into the grate; introduction of a reducing gas into the grate; and removing vaporized radionuclides from the rust. The method of claim 1, wherein less than five (5) percent of the graphite is gasified. The method of any preceding claim, wherein the radionuclides comprise carbon-14, and at least seventy (70) percent of the carbon-14 present is removed from the graphite. A method according to any preceding claim, wherein the radionuclides comprise carbon-14, and at least ninety (90) percent of the carbon-14 present is removed from the graphite. A process according to any one of the preceding claims, wherein the inert gas comprises at least one of nitrogen, helium, and argon and the reducing gas comprises at least one of hydrogen, hydrazine, ammonia, carbon monoxide and hydrocarbon vapor. A process according to any one of the preceding claims, wherein the reducing gas comprises one or more reducing gases which can give free hydrogen, carbon monoxide (CO), ammonium, or organic steam. A method according to any one of the preceding claims, further comprising the step of adding oxidizing gas to the rust. The method of claim 7, wherein the oxidizing gas comprises at least one of steam, carbon dioxide (CO 2), nitric oxide (N 2 O), oxygen (O 2), air, alcohols (with OH groups), or other oxygenated vapors. A method according to any one of the preceding claims, wherein the step of introducing the inert gas into the grate and introducing the reducing gas into the grate comprises introducing the inert gas and the reducing gas at a location near the bottom of the reactor and wherein the inert gas and the reducing gas flows through the graphite. A method according to any one of the preceding claims, further comprising the step of reducing the size of the graphite before introducing the graphite into the rust. A method according to any one of the preceding claims, wherein the grate comprises a vertically oriented moving bed reactor, and wherein the step of introducing graphite contaminated with radionuclides into the grate comprises the introduction of graphite near the top of the grate, and wherein the step of introducing the inert gas into the grate and the introduction of the reducing gas into the grate comprises the introduction of gases near the bottom of the grate. 10
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    引用文献:
    公开号 | 申请日 | 公开日 | 申请人 | 专利标题
    DE60024306T2|1999-10-14|2006-07-27|David Bradbury|METHOD FOR TREATING RADIOACTIVE GRAPHITE|
    US20080181835A1|2006-12-01|2008-07-31|Mason J Bradley|Steam reforming process system for graphite destruction and capture of radionuclides|
    FR2943167B1|2009-03-11|2011-03-25|Electricite De France|TREATMENT OF CARBON RADIOACTIVE WASTE.|RU2660804C1|2017-07-03|2018-07-10|Российская Федерация, от лица которой выступает Государственная корпорация по атомной энергии "Росатом"|Method of preparation of graphite radioactive waste to burial|
    RU2711292C1|2018-11-21|2020-01-16|Акционерное Общество "Российский Концерн По Производству Электрической И Тепловой Энергии На Атомных Станциях" |Nuclear reactor design decontamination method|
    法律状态:
    2019-12-03| NUG| Patent has lapsed|
    优先权:
    申请号 | 申请日 | 专利标题
    US13/278,786|US9040014B2|2011-10-21|2011-10-21|Graphite thermal decontamination with reducing gases|
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