NL8200332A - METHOD FOR ENCLOSING RADIO-ACTIVALLY POLLUTED OR RADIO-ACTIVE SUBSTANCES OF NUCLEAR DEVICES WITH A MATRIX SUITABLE FOR END-STORAGE. - Google Patents

Description

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Method for encapsulating radioactively contaminated or radioactive substances containing solids from nuclear installations with a .., ·. matrix suitable for final storage.

The invention relates to a method for encapsulating solids from radioactive pollutants or radioactive substances from nuclear installations with a matrix suitable for final storage.

Nuclear power plants, reprocessing factories and other nuclear facilities, in addition to weakly radioactive solid waste, also generate large amounts of medium and highly radioactive solid waste. The transfer of this waste in a state suitable for final storage has hitherto been effected in most cases by cementation in 200-liter drums or by encapsulating the solid waste with a cement milk to be hardened to cement stone.

When reprocessing sprayed fuel elements according to the Purex process, the fuel elements are treated with hot nitric acid and the nuclear fuels are largely dissolved.

The fuel element sleeves then remain undissolved and are separated from the fuel solution.

In addition to the dissolved fissile materials and fissile products, the fuel solution also contains undissolved solids of different composition and nature. The undissolved residues can lead to faults and blockages in the subsequent process steps due to deposition in the pipelines, transport and dosing systems. Especially in the subsequent extraction cycle, these undissolved solids can be accumulated in the boundary layers of the extraction and interfere very severely with the separation of the extraction vessels from each other.

In order to avoid such disturbances, a separation of the undissolved solution residue, the so-called "feed" settling sludge (FBS), from the nitric acid fuel solution is necessary.

8200332 t - 2 -

Essentially, two separation processes are eligible: a) centrifuge b) filtration

The process practiced in the Wiederaufarbeitungsanlage für bestrahlte Kernbrennstoffe Karlsruhe (WAK) has hitherto been considered as the only process for transferring feed settling sludge into a product suitable for final storage.

There the insoluble residue is separated from the fuel solution by plastic filter pouches and the filter pouches are reinforced with cement slurry together with sleeve and structural parts.

In a weathering facility (WA) with a larger processing amount, the solution residue must be separated from the fuel solution by centrifuge. The feed sedimentation sludge is then rinsed from the centrifuge with a weakly nitric acid solution.

This produces an FBS suspension of about 20 g / liter solids content. For the further treatment, a higher solids slurry without further process step would be desirable to obtain smaller waste volumes.

In the homogeneous cementation, between 160 to 170 liters of FBS suspension together with cement are mixed and stiffened in a 400 liter vessel. With a 350 t WA one has to deal with approximately. 440 barrels of stiffened feed sedimentation sludge.

The advantages of this known method lie in the simplicity and availability.

30 The main disadvantages of the heterogeneous and homogeneous FBS stiffening with cement according to this concept are the following: - large volumes of cemented waste - large activity inventory in relatively poor stiffening matrix 35 - radiation and heat load of the matrix - release of water from the matrix of radiolysis gas (including formation of hydrogen).

In France, England, and U.S.A. 40 feed sedimentation sludges or alone or with shells resp. other ____________ Λ 8200332 * M 'i - 3 - waste stored together under water interim. A method for stiffening the waste is not yet practiced there. The situation is similar for the fuel sleeves. At the WAK, the 5 fuel element sleeves are rolled and covered with cement slurry and thus made suitable for final storage. In other countries, as well as feed sedimentation sludge under water, they are stored for an indefinite period of time.

Cementation is also provided for other medium to highly radioactive solid waste.

All these products have the above-mentioned drawbacks.

The object of the invention is therefore based on the provision of a method for encapsulating solids from nuclear plants with radioactively contaminated or radioactive substances with nuclear matrix devices with a matrix suitable for final storage, wherein the disadvantages of the hitherto known processes and stiffening products are avoided. with substantially improved properties. Both the leaching resistance and the radiation and heat resistance of the stiffening products must be substantially increased compared to the corresponding properties of the prior art stiffening products. Release of radiolysis gas should be avoided. The process should be able to solidify medium to highly radioactive solid waste materials of any kind, as well as feed sedimentation sludge or fuel sleeve segments.

This object is achieved according to the invention in that the radioactive solids with the addition of glass frit in an amount containing 30 to 90 wt. %, based on the finished product, correspondingly, are introduced into an end storage container, in a sintering furnace 35 during a time period between 15 min and 50 hours at a pressure range between 20 and 500 bar and a temperature in the range 650 ° K to 950 ° K are sintered into a solid, and the end storage container is hermetically sealed thereafter.

40 This special purpose, namely the solution 8200332 - 4 - 4 v residue resp. To stiffen the leached fuel and / or culture element element sleeve segments according to the invention is solved in that c) the solution residue, i.e. the so-called.

5 feed sedimentation sludge (FBS), before or after separation of the fuel and / or culture medium solution, the sleeve segments after rolling, with glass frit in a ratio of FBS or ". sleeve segments to glass frit in the range of 10 wt. % at 90 wt. % to 70 wt. % at 30 wt. % 10 mixed resp. (d) the mixture is introduced into an end storage container, any remaining free space between mixture and container wall, respectively. container bottom is filled with glass frit, e) the loaded but open container is heated to a temperature in the range of 350 ° K to 523 ° K for up to 10 hours to remove moisture and other volatiles, and finally F) after transferring the container into a sintering furnace and after covering the mixture in the container, the contents of the container are kept for 15 minutes to 50 hours at a pressure between 20 and 500 bar and a temperature in the range of 650 ° K is sintered to 950 ° K.

Here, the FBS can be mixed with glass frit, or separately therefrom, before filtering with a filter medium, but at the same time added to the filter medium. The FBS is introduced jointly with the filtering agent into the final storage container. In order to reduce the fissile product losses, the FBS can be washed with dilute HNO4 after the filtering or centrifuging operation and before drying.

The mixture of radioactive solids 35 resp. of FBS and / or sleeve segments and glass frit can be covered in the end storage container with at least one inactive layer, such as, for example, with a layer of glass frit and / or a layer of a metal powder (for better heat conduction).

In the following the invention will be further elucidated on the basis of three diagrams for carrying out the method according to the invention, wherein scheme 1 shows the method for the encapsulation and stiffening of radioactive solids. -5 debris, reactor heart parts, structural parts or radioactive ceramic pellets, etc., scheme 2 illustrates the method of transferring FBS into a product suitable for final storage, and scheme 3 illustrates the method of transferring fuel and / or 10 culture element sleeves in a final storage facility appropriate product.

Method according to scheme 1:

According to the method of the invention, radioactive solids, which may consist, for example, of solid debris, reactor heart parts, structural parts, or ceramic pellets, etc., are placed together with glass frit in a scallop usable for final storage, and then placed in a sine sintered under pressure sintered ..

Process according to scheme 2: 20 According to the method of the invention, the slurry leaving the settling centrifuge (or the feed solution coming directly from the solubilizer) is fed together with a glass frit into a filter oven, in which the FBS and glass frit are simultaneously placed on the filter. wall is knocked down. Either the filter technique used in the WAK with plastic pouches or suitable filter cartridges made of sintered glass can be used as a filter.

The in proportion with FBS and glass frit (e.g.

20 wt. % FBS and 80 wt. Filter units filled with glass frit 30 may be washed again with dilute nitric acid to reduce fuel product losses. The suspension filtrate can be recycled back from the centrifuge to wash out the FBS. No further secondary waste 35 occurs in this way.

After washing and drying, the filter units are introduced into a steel mold, filling the bottom and further free space between the filter units and the mold wall with glass frit.

40 After the drying process previously performed, 8200332 - 6 - in which moisture and decomposition products of the plastic filter pouches, if used, escape from the mold, the contents of the mold are first covered with a layer of glass frit and if necessary pre-compacted by infiltration . Subsequently, a corresponding thick layer of a metal powder is applied to the glass frit and then the scallop in the inductively heated oven is brought under pressure to the desired sintering temperature of the glass frit according to a set program. After maintaining the sintering time (e.g. 1 to 3 hours) and a pressure of between 20 and 500 bar on the sintered product, the scallop is hermetically sealed at the same time by sintering the metal powder layer. To be on the safe side, a lid 15 with gripper button can be welded on.

The following advantages can be added to the processes that have been applied and are under development to date:

- small fuel losses due to the possibility of washing the FBS

- filtration in a flash vessel, which can be used directly in the final storage scallop

- high concentration of the FBS

- homogeneous distribution of FBS in the glass frit 25 - glass frit serves as a filter aid, since it is applied to the filter wall at the same time - no waste gas problems, since the sintering process takes place almost in closed mold - compared to other processes few 30 process steps - hydrogen-free matrix and therefore no radiolysis problems with the product - a high volume reduction factor compared to cementation.

35 Process according to scheme 3: Leached fuel element sleeves are dried with hot air after rinsing with nitric acid, changing their position quite often by shaking in order to remove all moisture.

40 Thereafter, the sleeves are rolled or rolled together with glass frit in the final storage mold, unrolled or rolled. * The filling is done in such a way that the closest possible packing of sleeves and glass frit is obtained before the sintering process. This can be achieved by corresponding shaking resp. infibrate.

- conclusions - 8200332

Claims (5)

Method for encapsulating solids from nuclear installations containing radioactively contaminated or radioactive substances with a matrix suitable for final storage, characterized in that the radioactive solids are added in an amount which, with the addition of glass frit, is added. corresponds to 30 to 90 wt. %, based on the final product, are introduced into an end storage container, in a sintering furnace for a period of time between 15 min and 50 hours at 10, a pressure in the range between 20 and 200 bar and a temperature in the range from 650 ° K to 950 ° K are sintered to a solid, and the end storage container is hermetically sealed thereafter. 2. Method for enveloping stiffening of nuclear fuel and / or culture element elements irradiated during reprocessing during the dissolving step in hot nitric acid, radioactive, undissolved solution residue and / or leached fuel and / or culture element element segments according to claim 1, wherein a) in the case of the solution residue, this is separated from the fuel and / or culture substance solution by filtration or by centrifugation, in the case of the sleeve segments these after the core fuel and / or culture matter have been dissolved therefrom and after separated from the solution by rolling them, and b) then mixing the solution residue and / or the sleeve segments with a stiffening matrix, respectively. is enveloped and transferred in a solid state suitable for final storage, characterized in that c) the solution residue, ie the so-called feed settling sludge (FBS), before or after the separation of the splitting and / or culture medium solution, the sleeve segments after rolling, with glass frit in a ratio of FBS 35 resp. sleeve segments to glass frit in the range of 10 wt. I at 90 wt. % to 70 wt. % at 30 wt. % mixed, resp. is enclosed, 8200332 - 9 - d) the mixture is introduced into an end storage container, any remaining space between the mixture and the container wall, respectively. container bottom is filled with glass frit, 5 e) the charged, still open container is heated for up to 10 hours to remove moisture and other volatiles at a temperature in the range of 350 ° K to 523 ° K, and finally f) after transferring the container into a sintering furnace and after covering the mixture in the container, the container contents for 15 minutes to 50 hours at a pressure in the range between 20 and 500 bar and a temperature in the range 650 ° K to 950 ° K is sintered under pressure. 3. A method according to claim 2, characterized in that the FBS is mixed with glass frit before filtering or mixed with a filter medium, or, however, it is fed simultaneously to the filter medium. 4. Method according to claim 2, characterized in that the FBS is introduced together with the filtering agent into the end storage container. 5. Process according to claim 2, characterized in that the FBS is washed with dilute HNO 2 after the filtering or centrifuging step and before drying with a view to reducing fuel product losses. / 8200332