GB2099207A

GB2099207A - Process for encasing radioactively contaminated solid substances or solid substances containing radioactive substance from nuclear plants in a matrix suitable for permanent storage

Info

Publication number: GB2099207A
Application number: GB8206958A
Authority: GB
Original assignee: Kernforschungszentrum Karlsruhe GmbH
Current assignee: Forschungszentrum Karlsruhe GmbH
Priority date: 1981-03-17
Filing date: 1982-03-10
Publication date: 1982-12-01
Also published as: DE3110192A1; BE892371A; GB2099207B; NL8200332A; JPS57163900A; FR2502381A1

Abstract

Radioactively contaminated solids or solids containing radioactive substances from nuclear plants are encased in a matrix suitable for permanent storage by producing a mixture of radioactive solids and glass frit in a container suitable for final storage and pressure sintering the mixture. The glass frit is introduced in a quantity corresponding to 30 to 90% by weight with respect to the weight of the mixture. The mixture in the container is pressure sintered in a sintering furnace for a period between 15 minutes and 50 hours at a pressure in the range between 20 and 500 bar and at a temperature in the range from 650 DEG K to 950 DEG K to form a solid body.

Description

SPECIFICATION Process for encasing radioactively contaminated solid substances or solid substances containing radioactive substances from nuclear plants in a matrix suitable for permanent storage The present invention relates to a process for encasing radioactively contaminated solid substances or solid substances containing radioactive substances from nuclear plants in a matrix suitable for permanent storage.

In nuclear power plants, reprocessing plants and other nuclear plants, large quantities of medium and highly radioactive solid wastes are produced in addition to weakly radioactive solid wastes. The transformation of these wastes into a state suitable for permanent storage has, in the past, in most cases, been effected by cementing them in 200 1 barrels or by encasing the solid wastes in a cement milk which hardens to form cement block.

During reprocessing of spent nuclear fuel elements according to the Pyrex process, the fuel elements are treated with hot nitric acid to substantially dissolve the nuclear fuels contained in the fuel elements into the nitric acid to thereby form a nitric acid fuel solution. The fuel element casings, however, remain undissolved and are separated from the fuel solution. The fuel solution contains, in addition to dissolved fission materials and fission products, undissolved solids of different composition and consistency. These undissolved solids or residues can form deposits in the pipelines and in the conveying and dosaging systems, thus causing malfunctions and stoppages during the subsequent process steps.

During the subsequent extraction-cycle, these undissolved solids can accumulate in the interfaces of the extraction phases, and have a very annoying effect on the separation of the individual extraction phases.

To avoid such malfunctions, it is necessary to separate the undissolved dissolver residue, the socalled feed clarification sludge (FKS), from the nitric acid fuel solution. Essentially two separation processes have been used for this purpose, namely, (a) centrifuging and (b) filtering.

So far, the process employed in the reprocessing plant for irradiated nuclear fuels at Karlsruhe (WAK) is the only process that can be considered to be suitable for converting feed clarification sludges into a product suitable for permanent storage. In this process, the undissolvable residue is separated from the fuel solution by passing the fuel solution through plastic filter bags. The filter bags, which now contain the undissolvable residue, are then solidified in a cement slurry together with the casings and structural components. This cementation process is referred to as a heterogeneous solidification.

In reprocessing plants with a largerthroughput than the above referred to Karlsruhe (WAK) plant, the dissolver residue is separated from the fuel solution by means of a centrifuge. In this process, the feed clarification sludge is washed out of the centrifuge with a weakly nitric acid washing solution.

This results in an FKS suspension having a solids content of about 20 g per liter. For further treatment, a suspension with a higher solids content which did not require further process steps would be desirable in orderto keep waste volume low.

For homogeneous cementation, between 160 and 170 liters of FKS suspension are mixed together with a cement and solidified in a 400 liter barrel. A 350 ton reprocessing plant would thus produce about 440 barrels of solidified feed clarification sludge per year.

The advantages of this known process are its simplicity and availability.

The heterogeneous and homogeneous FKS solidification processes with cement as described above have a number of major drawbacks. Thus, these solidification process produce large volumes of cemented wastes, and large amounts of activity are contained in a relatively poor solidification matrix. Further, the matrix is subject to radiation and heat stresses. The matrix contains water and thus releases radiolysis gases, for example hydrogen which is formed by radiolysis.

In France, England and the United States of America, feed clarification sludges are intermediately stored under water either by themselves or together with casings and other wastes.

A method for the solidification of wastes has not yet been put in practice in France, England and the United States. The situation is similar with respect to the fuel element casings. At WAK, the fuel element cases are rolled and covered with a cement slurry to thus be made suitable for permanent storage. In other countries, they are intermediately stored under water for an indetermined period of time similarly to the feed clarification sludges.

For other medium to highly radioactive solid wastes, cementation is likewise provided.

All these products have the above named drawbacks.

It is therefore an object of the present invention to provide a process for encasing radioactively contaminated solids or solids containing radioactive substances from nuclear plants in a matrix suitable for permanent storage, which process avoids the drawbacks of the prior art processes and produces solidification products having substantially improved properties.

A further object of the present invention is to provide such a process in which the leaching resistance, as well as the radiation and heat resistance, of the solidification products are significantly improved compared to the prior art solidification products.

A still further object of the present invention is to provide such a process which avoids release of radiolysis gas.

Another object of the present invention is to provide such a process in which it is possible to solidify medium to highly radioactive solid wastes of any type, as well as feed clarification sludges or fuel element casings.

Additional objects and advantages of the present invention will be set forth in part in the description which follows and in part will be obvious from the description or can be learned by practice of the invention. The objects and advantages are achieved by means of the processes, instrumentalities and combinations particularly pointed out in the appended claims.

To achieve the foregoing objects and in accordance with its purpose, the present invention provides a process for encasing radioactively contaminated solids or soldis containing radioactivate substances from nuclear plants in a matrix suitable for permanent storage, comprising introducing the radioactive solids and glass frit into a permanent storage container to provide a mixture of radioactive solids and glass frit, the glass frit being introduced in a quantity corresponding to 30 to 90 percent by weight with respect to the weight of the mixture, and pressure sintering the mixture in a sintering furnace for a period of time between 15 minutes and 50 hours at a pressure in the range between 20 and 500 bar and a temperature in the range from 650"K to 950"K to form a solid body.

It is to be understood that both the foregoing general description and the following detailed description are exemplary, but are not restrictive of the invention.

Figure 1 shows a process sequence in accordance with the present invention for encasing and solidifying broken radioactive pieces of solid substances, core members, structural components or radioactive ceramic pellets, etc.

Figure 2 shows a process sequence in accordance with the present invention for converting FKS into a product suitable for permanent storage.

Figure 3 shows a process sequence in accordance with the present invention for converting fuel andlor breeder element casings into a product suitable for final storage.

In the practice ofthe present invention, radioactively contaminated solids or solids containing radioactive substances from nuclear plants are encased in a matrix suitable for permanent storage.

The radioactive solids which can be treated in accordance with the present invention are highly active and medium active radioactive solids and can be, for example, broken pieces of a solid substance, core components, structural components, ceramic pellets, undissolved dissolver residue, fuel casing sections of breeder element casing sections (cladding material).

The present invention is based on the discovery that glass frit can be used to provide a matrix suitable for permanent storage upon sintering of a mixture containing the glass frit and radioactive solids.

The radioactive solids and glass frit are introduced into a container suitable for finial storageto provide a mixture of radioactive solids and glass frit. The glass frit is introduced in a quantity corresponding to 30 to 90 percent by weight with respect to the weight of the mixture of radioactive solids and glass frit. The mixture is then pressure sintered in a sintering fur nace for a period between 15 minutes and 50 hours at a pressure in the range between 20 and 500 bar and at a temperature in the range from 650"K to 950"K to form a solid body.

The present invention is especially concerned with the task of solidifying radioactive dissolver residue (FKS). The dissolver residue is the undissolved solids obtained during the reprocessing of irradiated nuclear fuel elements and/or breeder elements during the dissolution ofthese elements in hot nitric acid.

To achieve such a solidification of dissolver residue, the dissolver residue, i.e. the so-called feed clarification sludge (FKS), is mixed with or encased in glass frit in a weight ratio of FKS to glass frit of 10 percent by weight FKS to 90 percent by weight glass frit up to 70 percent by weight FKS to 30 percent by weight glass frit. The dissolver residue can be mixed with the glass frit before the dissolver residue has been separated from the nitric acid fuel solution and/or breeder material solution by filtering or centrifuging. Alternatively, the dissolver residue can be mixed with the glass frit after it has been separated, for example, by centrifuging, from the nitric acid fuel solution and/or breeder solution or can be mixed with the glass frit simultaneously with its separation from the nitric acid fuel solution and/or breeder solution.

After forming the mixture of dissolver residue and glass frit, the mixture is introduced into a permanent storage container, and any free spaces remaining between the mixture and the container wall or container bottom are filled with glass frit. The filled, still open container is then heated in the sintering furnace to a temperature in the range from 350"K to 523 K for a period of 5 to 10 hours so as to remove moisture and other volatile substances. Finally, after transferring the container into a sintering furnace and after covering the mixture in the container, the container contents are pressure sintered for 15 minutes to 50 hours at a pressure in the range between 20 and 500 bar and at a temperature in the range from 650"K to 950"K.

The mixture of dissolver residue (FKS) and glass frit preferably is introduced into a filtering medium, such as a filtering bag or a filter candle. The FKS can be mixed with glass frit before being introduced into the filtering medium, and then the mixture of FKS and glass frit can be introduced into the filtering medium and filtered therein. Alternatively, the FKS can be added to the filter separately from the glass frit, but simultaneously with the glass frit which acts as a filtering aid in the filtering medium. When adding the FKS and glass frit simultaneously to the filtering medium, the FKS can be in the form of a nitric acid fuel solution of Breeder solution from which the FKS has not been separated or can be an FKS which has been separated from the nitric acid solution as by centrifuging. The centrifuged FKS which is mixed with the glass frit preferably is in the form of a Suspension.

In order to reduce fission product losses, the FKS can be washed in the filtering medium with diluted HNO3 after the filtering and before drying. In addition, when centrifuging is used to separate the FKS from the nitric acid fuel solution and/or breeder material solution, the FKS can be washed in the centrifuge with diluted HNO3 after the centrifuging process and before drying.

The present invention is also especially concerned with the task of solidifying the leached fuel element and/or breeding element casing sections (cladding material). The casing sections are mixed with or encased in glass frit in a ratio of casing sections to glass frit of 10% by weight casing section to 90% by weight of glass frit up to 70% by weight casing sec tion to 30% by weight glass frit. The casing sections can be rolled to reduce their size before being mixed with the glass frit or can be mixed with the glass frit in an unrolled form.The mixture of casing sections and glass frit preferably is formed directly in the permanent storage container, and preferably care is taken to densely pack the mixture into the container, as by vibrating the container, so that there are no free spaces remaining between the mixture and the container wall or container bottom. Finally, after transferring the container into a sintering furnace and after covering the mixture in the container, the container contents are pressure sintered for 15 minutes to 50 hours at a pressure in the range between 20 and 500 bar and at a temperature in the range from 6500to 9500K.

The mixture of radioactive solids or of FKS or of casing sections and glass frit can be covered in the permanent storage container with at least one inactive layer, as, for example, with a layer of glass frit and/or a layer of a metal powder, in a layerthickness of 2 or 3 cm or more, for better heat dissipation. The metal powder can be a powder of aluminium, or aluminium alloys, or iron. During sintering ofthe mixture of radioactive solids and glass frit, the inactive layer, especially the layer of metal powder, simultaneously tightly seals the permanent storage container by sintering. After sintering, a cover can be welded onto the permanent storage container to tightly seal it.

Referring now to Figure 1, radioactive waste materials which can be highly active wastes (HAW) and medium active wastes (MAW) and can comprise, for example, broken pieces of a solid substances 10, core components 12, structural components 14 or ceramic pellets 16, or other highly active and medium active wastes etc., are introduced, as represented by arrow 18, into a mold 20 usable for permanent storage. The radioactive waste materials 18 are introduced into mold 20 together with glass frit, as represented by arrows 22.

In introducing the radioactive waste materials and glass frit into mold 20, preferably care is taken to densely pack the waste materials and glass frit into the container, as by vibrating the container, so that there are no free spaces remaining between the resulting mixture and the mold wall and mold bottom. Generally, the weight ratio of radioactive waste material to glass frit employed in filling mold 20 is 10:90 to 70:30.

Mold 20 is then transferred to an inductively heated sintering furnace 24. The content of mold 20 is initially covered with a layer ofglassfrit 26 and, if necessary, is precompressed somewhat by vibration. Thereafter, a correspondingly thick layer of a metal powder 28 is piled onto glass frit layer 26 (after introducing the mold in the sintering furnace) and then, under pressure and according to a given program, mold 20 is brought to the desired sintering temperature of the glass frit so as to sinter the glass frit in the inductively heated sintering furnace 24.

While maintaining the sintering time (e.g. 1 to 3 hours) and a pressure between 20 and 500 bar on the sinter product, mold 20 is simultaneously sealed tightly by sintering the layer of metal powder 28. As a safety measure, a cover 30 equipped with a grip ping head may ultimately be welded on (after sinter ing).

Turning now to Figure 2, there is shown a filter 32 which is fed with dissolver residue through a conduit 34 and/or a conduit 36. The dissolver residue which is fed into filter 32 through conduit 34 is in the form of a nitric acid feed solution, that is, a nitric acid nuclear fuel solution and/or breeder solution, as it comes directly from the dissolver. The dissolver residue which is fed into filter 32 through line 36 is in the form of an FKS suspension which is formed in a clarification centrifuge 38. Centrifuge 38 is fed with a nitric acid feed solution as it comes directly from the dissolver through a conduit 40, and separates the undissolved dissolver residue from the solution to form an FKS suspension which can contain, for example, 20g of solids per liter.The dissolver residue (FKS) coming from conduits 34 and/or 36, together with the glass frit, as represented by arrow 42, are transferred into filter 32 in which the FKS and glass frit are simultaneously precipitated on the filter wall. Thus, there is a filtering of the dissolver residue simultaneously with the addition of glass frit. Filter 32 can be of the type including a filter unit in the form of a plastic bag as represented by reference numeral 44 as employed in the filtering technique of WAK, or it may include a filter unit in the form of a suitable filter candle of sintered glass, as represented by reference numeral 46. The filter units, either bags 44 and/or candles 46, are filled with FKS and glass frit in the desired weight ratio, for example, 20 percent by weight FKS and 80 percent by weight glass frit.After filling, each filter unit can be washed once more with diluted nitric acid so as to reduce fission product losses. The filtrate 48 of this washing can again be used to wash the FKS out of centrifuge 38. In this way, there is no secondary waste.

After this washing of the filter units, the filled filter units 44 and 46 are inserted into a steel mold 50 together with glass frit, as represented by arrow 52, so that there is a simultaneous filling of mold 50 with the bottom and other free spaces between the filter units and the mold wall being filled with glass frit.

Mold 50 is then transferred to an inductively heated sintering furnace 54.

Afterthe drying process, at for example 250"C in the furnace, in which moisture and decomposition products from plastic bags 44 escape from mold 50, if such bags are used, the content of mold 50 is initially covered with a layer 56 of glass frit and, if necessary, is precompressed somewhat by vibration. Thereafter, a correspondingly thick layer of a metal powder 58 is piled onto glass frit layer 56 and then, under pressure and according to a given program, mold 50 is brought to the desired sintering temperature of the glass frit so as to sinter the glass frit in the inductively heated sintering furnace. While maintaining the sintering time (e.g. 1 to 3 hours) and a pressure between 20 and 500 bar on the sinter product, mold 50 is simultaneously sealed tightly by sintering the layer of metal powder 58.As a safety measure, a cover 60 equipped with a gripping head may ultimately be welded on.

The process of the present invention for heating dissolver residue has a number of advantages as compared to the presently practiced processes and those in the development stage. Thus, the process of the present invention results in lower fuel losses because of the possibility of washing the FKS in centrifuge 38 and filter 32. Further, in the process of the present invention, there is a filtration into a filtering vessel (bags 44 or candle 46) which can be inserted directly into the mold used for final storage. The process of the present invention enables the use of high concentrations of FKS (for example 20 to 30% by weight) and provides for a homogeneous distribution of FKS in the glass frit. In addition, the glass frit serves as a filtering aid since it is simultaneously applied to the filter wall with the FKS.No exhaust gas problems arise since the sintering process takes place practically in a closed mold. There are a smaller number of process steps compared to prior art processes. The matrix is free of hydrogen and thus there are no radiolysis problems in the product.

Finally, there is a high volume reduction factor of ca.

50 compared to cementation.

Referring now to Figure 3, there is shown leached fuel element casings or leached breeder element casings in the form of dissolver baskets 62 which are dried with hot air after having been rinsed with nitric acid. During the drying, their position is frequently changed by shaking, so as to remove all moisture.

Then casings 62 are filled into a permanent stor age mold 64 together with glass frit 66. Casings 62 can be rolled, as by rollers 68, on their way into mold 64 as shown in scheme "a", or can be filled into mold 64 in an unrolled form as shown in scheme "b". In either event, the filling is effected in such a manner that casings 62,69 and glass frit 66 are packed as densely as possible even before the sintering process. This dense packing can be obtained by appropriately shaking or vibrating the material in mold 64. Mold 64 preferably is filled so that it has a top layer of glass frit.

Mold 64 is then transferred to an inductively heated sintering furnace 70. A thick layer of a metal powder 72 is placed on the top layer of glass frit and then, under pressure and according to a given program, mold 64 is brought to the desired sintering temperature of the glass frit so as to sinterthe glass frit in the inductively heated furnace. While maintaining the sintering time (e.g. 1 to 3 hours) and a pressure between 20 and 500 bar on the sinter product, mold 64 is simultaneously sealed tightly by sintering the layer of metal powder 72. As a safety measure, a cover 74 equipped with a gripping head may ultimately be welded on.

Claims

1. Process for encasing radioactively contaminated solids of solids containing radioactive substances from nuclear plants in a matrix suitable for permanent storage, comprising: a) introducing the radioactive solids and glass frit into a container suitable forfinal storage to provide a mixture of radioactive solids and glass frit, the glass frit being added in a quantity corresponding to 30 to 90 percent by weight with respect to the weight of the mixture, and b) pressure sintering the mixture in a sintering furnace for a period between 15 minutes and 50 hours at a pressure in the range between 20 and 500 bar and at a temperature in the range from 650"K to 950"K to form a solid body.

2. Process according to claim 7, wherein the radioactive solids which are mixed with the glass are radioactive undissolved residue obtained during the reprocessing ofirradiated nuclear fuel and/or breeder elements during dissolution in hot nitric acid to form a fuel solution or breeder solution, and wherein:: the dissolver residue is separated from the fuel and/or breeder solution by filtering or centrifuging; the dissolver residue and glass frit are formed into said mixture before they are introduced into the container suitable forpermanent storage; the mixture is introduced into the container suitable for permanent storage and any remaining free spaces between the mixture and the container wall or container bottom are filled with glass frit; the filled, still open container is heated to a temperature in the range from 350"K to 523"K for a period up to 10 hours so as to remove moisture and other volatile substances; and after the container has been transferred to a sintering furnace and the mixture in the container has been covered, the contents of the container are subjected to said pressure sintering.

3. Process as defined in claim 2, wherein the dissolver residue is separated from the nitric acid fuel and/or breeder solution before the dissolver residue is mixed with the glass frit.

4. Process as defined in claim 2, wherein the dissolver residue is separated from the nitric acid fuel and/or breeder solution after the dissolver residue has been mixed with the glass frit.

5. Process as defined in claim 2, wherein the dissolver residue is separated from the nitric acid fuel and/or breeder solution while the dissolver residue is being mixed with the glass frit.

6. Process as defined in claim 2, wherein the FKS is filtered with a filtering medium and is mixed with glass frit before being filtered with the filtering medium.

7. Process as defined in claim 2, wherein the FKS is added to a filtering medium separately from the glass frit but simultaneously with glass frit which acts as a filtering aid.

8. Process as defined in claim 2, wherein the FKS is introduced into the permanent storage container together with a filtering medium.

9. Process as defined in claim 2, wherein the FKS is filtered and the FKS is washed with diluted HNO3 after the filtering and before drying so as to reduce fission product losses.

10. Process as defined in claim 2, wherein the FKS is centrifuged and the FKS is washed with diluted HNO3 after the centrifuging and before dry ing so as to reduce fission product losses.

11. Process according to claim 1, wherein the radioactive solids are leached fuel and/or breeder element casing sections (cladding material).

12. Process according to claim 11, wherein said casing sections have been separated from a nuclear fuel and/or breeder material solution, and the separated casing sections are mixed with the glass frit.

13. Process according to claim 12, wherein the separated casing sections are rolled before being mixed with the glass frit.

14. Process according to claim 12 or 13, wherein the container is filled with the casing sections and glass frit, and the filled still open container is heated to a temperature in the range from 350"K to 523 K for a period up to 10 hours so as to remove moisture and other volatile substances; and after the container has been transferred to a sintering furnace and the mixture in the container has been covered, the contents of the container are subjected to said pressure sintering.

15. Process according to claim 1, wherein after pressure sintering, the permanent storage container is tightly sealed.

16. Process according to claim 2, wherein the mixture in the container is covered with a layer of metal powder.

17. Process according to claim 2, wherein the mixture in the container is covered with a layer of glass frit.

18. Process for encasing radioactively contaminated solids or solids containing radioactive substances from nuclear plants in a matrix suitable for permanent storage, substantially as hereinbefore described with reference to and as illustrated in the accompanying drawings.