EP0054604A2 - Process for preparing spent solid bodies for the final disposal of radioactive wastes - Google Patents

Abstract

1. Method for producing waste solidification products suitable for final storage and containing radioactive substances with a high radiation resistance and a low radiolytic gas generation, respectively, of the category of the molded radioactive waste bodies which are solidified with hydraulically binding, hardenable, inorganic materials, wherein radioactive, liquid, aqueous and/or radioactive solid wastes are mixed or coated with a cement-water suspension, are allowed to harden and are subjected to a heat treatment, comprising allowing hardening the mixtures of waste and cement-water suspension with a Portland cement klinker containing cement in closed containers at a temperature between room temperature up to 150 degrees C to become molded bodies and subjecting the hardened molded bodies to a heat treatment at normal pressure or negative pressure at temperatures of 100 degrees C up to 250 degrees C at the maximum so as to remove the water still unbound in the pores of the molded bodies.

Description

The invention relates to a process for the production of waste solidification products that are ready for disposal and contain radioactive materials with increased radiation resistance or reduced radiolysis gas formation from the category of the radioactive waste molded bodies solidified with hydraulically setting, hardening, inorganic materials, in which radioactive, liquid, aqueous and / or radioactive solid wastes are mixed or coated with a cement-water slurry, allowed to harden and subjected to a heat treatment.

• - hoch- und mittelradioaktive Stoffe und/oder Aktiniden enthaltende, wäßrige _Abfallkonzentrate,
• - in Wasser aufgeschlämmte, feinkörnige, feste Abfälle
• - aktinidenhaltige Verbrennungsrückstände,
• - feste, stückige Abfälle, wie Brennelementhülsen oder Core-Bauteile etc..

The radioactive, liquid, aqueous and / or radioactive solid wastes, which are solidified with a high degree of safety for the environment in the present process and incorporated in the hardening shaped bodies, are those such as, for example

• - high- and medium-level radioactive bfallkonzentrate substances and / or actinide-containing aqueous _A,
• - fine-grained, solid waste suspended in water
• - combustion residues containing actinides,
• - solid, lumpy waste, such as fuel element sleeves or core components etc.

For solidifying radioactive, aqueous waste, the waste concentrates is already long been proposed to reduce the volume of such wastes, thereby concentrate the radioactive material to undergo with an addition of glass formers to a heat treatment either to _'spread the radioactive substances in the resulting glass melt. -exist and then allow the melt to solidify to a solid (DE-OS 26 09 299), or to mix with silicate-containing clays or ion exchangers and to burn ceramic to a solid body (DE-PS 27 26 087).

The disadvantages of the production of glass blocks which contain radioactive waste materials can be seen in the fact that this requires a relatively complicated and costly process technology. Another disadvantage, both when solidifying in glass and in clays that are fired, is that in high-temperature steps from the as yet unconsolidated waste significant amounts of radioactive materials, such as e.g. Ruthenium, cesium etc., volatilize and have to be retained by complex exhaust gas cleaning with solid filters, condensate separators and washing columns.

Cement mixtures are currently used to solidify solid, lumpy waste such as fuel element sleeves or core components. The products are not post-treated thermally. The disadvantages of cemented high-level radioactive or medium-level radioactive waste, ash, fuel element sleeves and core component products are that, due to the comparatively high water content under the influence of the radioactive radiation emanating from the waste, large amounts of hydrogen and Oxygen are formed. This could be a disadvantage for the Bring operational safety to the interim and final storage. In addition, the temperatures of the products during storage must be regulated so that they do not exceed 100 ° C. At higher temperatures, water is released from the products, which leads to significant pressure build-up in the product itself and in the container. If the container is corroded, this can lead to the release of activity from the container.

The invention is based on the object of creating a process for the production of waste solidification products ready for disposal, containing radioactive materials, with increased radiation resistance or reduced radiolysis gas formation from the category of the radioactive waste molded bodies solidified with hydraulically setting, hardening, inorganic materials that all or nearly all the positive properties of the known solidification products also have, but the disadvantages of the known products, in particular the _R adiolysegas formation, the pressure buildup in the product or in the container, the risk of activity release from the container or the volatilization of radioactive Do not have nuclides in the production during the high temperature steps etc.

The object is achieved in a surprisingly simple manner according to the invention in that the mixtures of waste and cement-water slurry in closed containers at a temperature between room temperature can be allowed to harden to shaped bodies up to 150 ° C and that the hardened shaped bodies at a pressure between normal pressure and vacuum are subjected to a heat treatment at temperatures up to a maximum of 250 ° C to remove the unbound water present in the pores of the shaped bodies.

• a) der radioaktive Abfall mit einer Aufschlämmung eines portlandzementklinkerfreien Zementes vermischt bzw. umhüllt wird,
• b) die aus a) erhaltenen Mischungen in geschlossenen Behältern bei Raumtemperatur oder bei einer Temperatur oberhalb Raumtemperatur bis 150°C zu geformten Körpern erhärten gelassen werden,
• c) die erhärteten Formkörper bei einem Druck zwischen Normaldruck und Vakuum einer Wärmebehandlung bei Temperaturen bis maximal 250°C zur Entfernung des in den Poren der Formkörper ungebunden vorliegenden Wassers unterzogen werden.

An advantageous embodiment of the invention using special cements is characterized in that

• a) the radioactive waste is mixed or coated with a slurry of cement that is free of Portland cement clinker,
• b) the mixtures obtained from a) are allowed to harden into shaped bodies in closed containers at room temperature or at a temperature above room temperature up to 150 ° C.,
• c) the hardened moldings at a pressure between normal pressure and vacuum are subjected to a heat treatment at temperatures up to a maximum of 250 ° C. to remove the water present in the pores of the moldings.

According to the invention, the waste is mixed or coated with a cement which contains 20 to 30% by weight of SiO ₂ and 40 to 70% by weight of CaO, or 25 to 35% by weight of SiO ₂ , 10 to 25. wt .-% CaO and 30 to 5 ₀ wt .-% Al203 contains. The cement used can also contain additives such as bentonite. A advantageous embodiment of the invention is characterized in that the heat treatment of the shaped body (step c) a temperature between 150 ⁰ and 250 ⁰ C and is carried out over a period of between 12 hours and 4 days.

• 1. Hochaktive Abfallösungen -das sind salpetersaure Lösungen von vorwiegend Schwermetallnitraten-, die bei der Abtrennung der Spaltprodukte aus abgebrannten Kernbrennstoffen bei deren Wiederaufarbeitung entstehen.
• 2. Mittelaktive Abfallösungen -das sind überwiegend salpetersaure Lösungen mit in der Regel hohen Gehalten an Natriumnitrat, die an anderer Stelle als die hochaktiven _Abfallösungen bei der Wiederaufarbeitung von Kernbrennstoffen und bei der Dekontamination in kerntechnischen Anlagen anfallen.
• 3. Aktinidenkonzentrate, das sind Lösungen, Pulver, oder Verbrennungsrückstände, die vor allem bei der Verarbeitung und Fabrikation von Kernbrennstoffen als Abfallprodukte erhalten werden.
• 4. Aschen und Rückstände aus der Verbrennung organischer radioaktiver Abfälle.
• 5. Feste, stückige, metallische Abfälle wie Brennelementhülsen, Core-Bauteile, sonstige aktivierte und/oder kontaminierte metallische Abfälle, die in einem Betrieb von kerntechnischen Anlagen und bei der Wiederaufarbeitung von abgebrannten Kernbrennstoffen anfallen.

For example, in the reprocessing of nuclear fuels, in the manufacture and processing of nuclear fuels and in other nuclear facilities, the following radioactive waste is produced, which can be solidified using the method according to the invention:

• 1. Highly active waste solutions - these are nitric acid solutions of predominantly heavy metal nitrates - that arise when the fission products from spent nuclear fuels are separated and reprocessed.
• 2. Intermediate-level waste solutions -the are predominantly nitrate solutions with the generally high levels of sodium nitrate, which bfallösungen elsewhere than the highly active _A at the reprocessing of nuclear fuel and obtained in the decontamination of nuclear facilities.
• 3. Actinide concentrates are solutions, powders, or combustion residues that are obtained as waste products, especially in the processing and manufacture of nuclear fuels.
• 4. Ashes and residues from the incineration of organic radioactive waste.
• 5. Solid, lumpy, metallic waste such as fuel element cores, core components, other activated and / or contaminated metallic waste that occurs in the operation of nuclear facilities and in the reprocessing of spent nuclear fuels.

The resulting aqueous solutions or aqueous slurries are mixed directly with the cement according to the invention with stirring in a container, which may also be the final storage container, or mixed indirectly in a mixing vessel and then filled into the container. The solid, lumpy waste is poured over with a mixture made of cement according to the invention and water (W / Z value advantageously 0.3 to 0.6), possibly using low pressures. The hydration (process step b)) is advantageously carried out at temperatures between room temperature and 100 ° C. The removal of the excess water (process step c)) is advantageously carried out at temperatures between 100 ° C and 200 ° C. The temperature and the duration of the heat treatment are strongly dependent on the composition and dimensions of the products. The products should be heated up relatively slowly to avoid cracking. The optimal heating and drying times must be adapted to the respective product composition and size. Heating rates in the range between 0.5 C / min have proven to be advantageous. and 2 ° C / min. lie.

Compared to the previous known methods for solidifying the waste solutions or slurries, such as. ^B. solidification in a glass matrix, etc., the inventive method has a number of significant advantages.

For example, the process technology for producing the waste products is very simple compared to glazing (mixing in the container by stirring in a separate mixing vessel and filling into the container). The hydration of the products and the removal of the excess water (the water present in the pores of the solidification product) takes place at comparatively low temperatures, therefore no radioactive elements escaping via the gas phase, e.g. Cesium and ruthenium.

Due to the low temperatures during the product fabrication, the corrosion _rates' of the apparatus used to be very low.

By reducing the water content in the products through the thermal aftertreatment, the formation of radiolysis gas in the products is significantly reduced. The mechanical resistance is not affected by this treatment, since it is only heated to a comparatively low temperature and only excess water is separated, which is not involved in the structure of the cement stone.

By decreasing the water content of the products can be stored at temperatures which exceed 100 ⁰ C to 150 ° C, since no free water is no longer present and therefore can not build an elevated water vapor partial pressure in the container and no activity can be released.

When using special cements (e.g. deep drilling cements) products are obtained which are much more resistant to the attack of aqueous salt solutions and are also more thermally resistant than products which correspond to the state of the art.

• Durch die thermische Nachbehandlung bis ca. 250°C wird die thermodynamisch mögliche Reaktion zwischen Zirkon und Wasser schnell ablaufen und durch die entstehende Oxidschicht schnell zum Stillstand kommen. Aufgrund der thermischen Nachbehandlung steht dann kein reaktionsfähiges Wasser mehr zur Verfügung. Dies bedeutet, daß die bei Untersuchungen festgestellten hohen Tritiumfreisetzungsraten aus konventionellen Brennelementhülsen-Zementprodukten bei den Verfestigungsprodukten gemäß der Erfindung nicht mehr auftreten werden. Ähnliches gilt für zementierte, tritiumhaltige Core-Bauteile.

There is a further advantage with fuel bundle solidification products:

• Thanks to the thermal aftertreatment up to approx. 250 ° C, the thermodynamically possible reaction between zircon and water will take place quickly and will quickly come to a standstill due to the oxide layer. Due to the thermal aftertreatment, reactive water is then no longer available. This means that the high tritium release rates found in tests from conventional fuel element cement products will no longer occur in the solidification products according to the invention. The same applies to cemented, tritium-containing core components.

The invention is explained in more detail in the following examples without, however, being restricted to these examples.

Example 1:

• Es wurden simulierte Verfestigungsblöcke im Labormaßstab (ca. 85 mm hoch und ca. 62 mm Durchmesser) aus Portlandzement 35 (PZ 35) mit einem Wasser-Zement-Wert W/Z = 0,4 hergestellt und zwar
  • a) mit Leitungswasser (Produktbezeichnung PZ 35/Zementstein (ohne r.a. Abfallj)
  • b) mit simulierter mittelradioaktiver Abfallösung (Produktbezeichnung PZ 35/MAW)
  • c) mit simulierter hochaktiver Abfallösung (PZ 35/ HAW).

Reduction of radiolysis gas formation as a result of the heat treatment:

• Simulated consolidation blocks on a laboratory scale (approx. 85 mm high and approx. 62 mm diameter) were produced from Portland cement 35 (PZ 35) with a water / cement value W / Z = 0.4
  • a) with tap water (product designation PZ 35 / cement stone (without ra wastej)
  • b) with simulated medium-radioactive waste solution (product designation PZ 35 / MAW)
  • c) with simulated highly active waste solution (PZ 35 / HAW).

Half of the samples from a), b) and c) were produced in accordance with the prior art, ie without heat treatment, only stored for 28 days at room temperature. The other half of the samples were heat-treated according to the method of the invention, ie dewatered at 200 ° C. Then all samples were irradiated with 10 MeV electrons up to 10 ⁸ rad and the gas yield measured in ml / g and Mrad. The weight loss of the dewatered samples was 23% by weight both in the samples under a) and in the samples under b) and under c). This corresponds to a residual water content of about 7 to 9 _G ew .-% in the product. Radiolysis gas formation for all samples is shown in the table below.

It is clear from the measured values that the samples treated by the method according to the invention had less hydrogen generation by one to two powers of ten.

The table below shows the compressive strength of the sample blocks.

Compressive strength:

PZ 35 / cement stone (without ra waste): W / Z = 0.4

PZ 35 / MAW: W / Z = 0.4

PZ 35 / HAW:

W / Z = 0.4

The results show that the mechanical properties of the solidification products produced by the process according to the invention are not or only insignificantly deteriorated compared to the blocks produced according to the prior art.

Example 2:

• Wie in Beispiel 1 wurden Laborblöcke hergestellt, jedoch unter Verwendung eines Tiefbohrzementes,welcher keinen Portlandzementklinkeranteil aufweist und in die Gruppe Zemente zu rechnen ist, deren Hauptbestandteile in die Bereiche 20 bis 30 Gew.-% SiO₂ und 40 bis 70 Gew.-% Ca0 fallen, anstelle des PZ 35. Es wurden nur Proben entsprechend a) in Beispiel 1 hergestellt. Der Wasser-Zement-Wert betrug ebenfalls ₀,4, die Bestrahlungsart und -Dauer war die gleiche wie in Beispiel. 1. Die Wasserstoffbildungsrate betrug bei der bei Raumtemperatur 28 Tage lang gelagerten Probe (Stand der Technik) 2,5 . 10-3 ml/g und _Mrad und bei der nach dem erfindungsgemäßen Verfahren behandelten (bei 200°C entwässerten) Probe nur 3,7 10^-4 ml/g und Mrad.

Reduction of the radiolysis gas formation as a result of the heat treatment using a special cement:

• As in Example 1, laboratory blocks were produced, but using a deep-drilling cement, which has no Portland cement clinker content and is to be included in the group of cements, the main components of which are in the ranges 20 to 30% by weight SiO ₂ and 40 to 70% by weight Ca0 fall, instead of the PZ 35. Only samples corresponding to a) in Example 1 were produced. The water- Cement ratio was also _0, 4, and duration of irradiation was the same as in Example. 1. The rate of hydrogen formation in the sample stored at room temperature for 28 days (prior art) was 2.5. 10-3 ml / g and _M rad and only 3.7 10 ^-4 ml / g and Mrad in the sample treated (dewatered at 200 ° C.) according to the method of the invention.

Example 3:

• a) Zur Simulierung einer umhüllenden Verfestigung von festen stückigen radioaktiven Abfällen wie Brennelementhülsen, die bei der Wiederaufarbeitung bestrahlter Kernbrenn- und/oder Brutstoffelemente anfallen, oder wie Core-Bauteile, wurde eine Zement-Wasser-Mischung aus einem speziellen Zement, welcher keinen Portlandzementklinkeranteil aufweist und in die Gruppe Zemente zu rechnen ist, deren Hauptbestandteile in die Bereiche 25 bis 35 Gew.-% SiO₂, 10 bis 25 Gew.-% CaO und 30 bis 50 Gew.-% Al203 fallen, hergestellten und unter Anwendung von geringem Druck in eine Form vergossen und ohne Zusatz inaktiver metallischer Stücke erhärten gelassen. Eine derartige Simulierung ist zulässig, weil die metallischen Stücke nicht am chemischen Geschehen während der Erhärtung der Zement-Wasser-Mischung teilnehmen. Zusätze an Metallstücken würden lediglich die _Festigkeits-Eigenschaften verbessern, die jedoch zur Lösung der vorliegenden Aufgabe praktisch nichts beitragen. Die Zement-Wasser-Mischung hatte einen Wasserzementwert von W/Z = 0,45, der Probenkörper wurde 4 Tage lang bei 100°C ausgehärtet und danach 2 Tage lang bei ebenfall 100°C im Trockenschrank getrocknet. Es wurde danach keine Rißbildung beobachtet, der Gewichtsverlust betrug ca. 10%. Auf diese Weise hergestellte Proben wurden anschließend mit-γ-Strahlen bestrahlt und die durch Radiolyse erzeugte WasserstoffAusbeute der bestrahlten Zementsteinproben gemessen. Zum Vergleich wurden unter denselben Bedingungen Zementsteinproben aus Portlandzement 35 hergestellt und untersucht. Die Ergebnisse dieser Untersuchungen sind in den folgenden Tabellen aufgeführt und zum Vergleich einander gegenübergestellt.
  
• b) Laborblöckchen, die unter Verwendung eines speziellen Zementes (wie oben beschrieben) hergestellt wurden, erwiesen sich bei Korrosionsuntersuchungen während der Lagerung in quinärer Lösung (Salzlösung mit der Zusammensetzung in Gew.-%: 24,7 % MgCl₂: 2,3% MgSO₄; 1, 9%. NaCl ; 3,3% KC1; 67,9 H₂0) wesentlich beständiger als die Blöckchen aus herkömmlichem Portland-Zement, wie einerseits eine visuelle Beurteilung und anderseits auch klar die Massenänderungen der unbestrahlten Proben während der Lagerung in der quinären Lösung aufzeigten.
  

Comparative investigations on cement stone samples without simulated waste (a) Radiolysis hydrogen yield as a function of the pre-storage time in γ-irradiation; b) Weight changes when storing unirradiated samples in quinary solution at 100 ° C and a bar):

• a) In order to simulate an enveloping solidification of solid, lumpy radioactive waste such as fuel element sleeves, which result from the reprocessing of irradiated nuclear fuel and / or fuel elements, or like core components, a cement-water mixture was made from a special cement that does not contain any Portland cement clinker and the group includes cements, the main components of which fall in the ranges 25 to 35% by weight of SiO ₂ , 10 to 25% by weight of CaO and 30 to 50% by weight of Al 2 O 3, produced and using low pressure Poured into a mold and left to harden without the addition of inactive metallic pieces. Such a simulation is permissible because the metallic pieces do not participate in the chemical process during the hardening of the cement-water mixture to take. Additions of metal pieces would improve only the _F estigkeits properties, but which contribute to solving the present problem practically nothing. The cement-water mixture had a water cement value of W / Z = 0.45, the test specimen was cured at 100 ° C. for 4 days and then dried in a drying cabinet at 100 ° C. for 2 days. No cracking was then observed, the weight loss was approximately 10%. Samples produced in this way were then irradiated with γ-rays and the hydrogen yield generated by radiolysis of the irradiated cement stone samples was measured. For comparison, cement block samples were made from Portland cement 35 and examined under the same conditions. The results of these tests are shown in the following tables and compared for comparison.
  
• b) Laboratory blocks that were produced using a special cement (as described above) were found to be corrosion-proof during storage in quinary solution (saline solution with the composition in% by weight: 24.7% MgCl ₂ : 2.3% MgSO ₄ ; 1, 9%. NaCl; 3.3% KC1; 67.9 H ₂ 0) much more resistant than the blocks made of conventional Portland cement, such as a visual assessment on the one hand and on the other hand also clearly the changes in mass of the unirradiated samples during the Show storage in the quinary solution.
  

In Table 2, M means the weight, weighed after certain time intervals, for the corresponding sample, ΔM the weight change in percent of Starting weight. From the table it can easily be seen that after a certain period of time the weight change of the sample with the special cement becomes practically zero, ie after a certain amount of the quinary solution has been taken up, the mass of the sample remains the same. In contrast, the Portland cement sample (which corresponds to the prior art) swells, exhibits a significant increase in mass during increasing storage in the quinary solution and is therefore exposed to increased corrosion.

Example 4:

• Bei der Verfestigung von wäßrigen Konzentraten oder Aufschlämmungen (simulierter MAW) wird das Konzentrat mit dem Zement vermischt. Unter Verwendung eines Spezial-Zementes ohne PZ-Klinker (wie in Beispiel 3 beschrieben) wurden mit MAW-Simulat Proben mit einem Salzgehalt von 10 Gew.-% hergestellt (W/Z = 0,45). Bei Korrosionsuntersuchungen in quinärer Lösung bei 100⁰C erwiesen sich diese Produkte wesentlich korrosionsbeständiger als vergleichbare Produkte aus herkömmlichem Portland-Zement.
  

Comparative studies on changes in weight when storing cement / MAW samples in quinary solution:

• When aqueous concentrates or slurries (simulated MAW) solidify, the concentrate is mixed with the cement. Using a special cement without PZ clinker (as described in Example 3), samples with a salt content of 10% by weight were produced with MAW simulate (W / Z = 0.45). In corrosion tests in quinary solution at 100 ⁰ C, these products proved to be significantly more corrosion-resistant than comparable products made from conventional Portland cement.
  

Claims (5)

1. Process for the production of waste solidification products that are ready for disposal and contain radioactive materials with increased radiation resistance or reduced radiolysis gas formation from the category of the radioactive waste molded bodies solidified with hydraulically setting, hardening, inorganic materials, in which radioactive, liquid, aqueous and / or radioactive solid wastes are mixed or coated with a cement-water slurry, allowed to harden and subjected to a heat treatment,
characterized in that
the mixtures of waste and cement-water slurry are allowed to harden in closed containers at a temperature between room temperature to 150 ° C. to give shaped bodies and that the hardened shaped bodies are subjected to heat treatment at temperatures of up to 250 ° C. at a pressure between normal pressure and vacuum Removal of the unbound water present in the pores of the moldings are subjected. 2. Process for the production of _A bf _' all solidification products which are ready for disposal and contain radioactive substances and which have increased radiation resistance or reduced radiolysis gas formation from the category of radioactive solidified with hydraulically setting, hardening, inorganic materials Waste moldings in which radioactive, liquid, aqueous and / or radioactive solid wastes are mixed or coated with a cement-water slurry, allowed to harden and subjected to a heat treatment,
characterized in that a) the radioactive waste is mixed or coated with a slurry of cement that is free of Portland cement clinker, b) the mixtures obtained from a) are allowed to harden in molded containers at room temperature or at a temperature above room temperature to 150 ° C., c) the hardened moldings at a pressure between normal pressure and vacuum are subjected to a heat treatment at temperatures up to a maximum of 250 ° C. to remove the water present in the pores of the moldings. 3. The method according to claim 2, characterized in that the mixing or coating of the waste is carried out with a cement containing 20 to 30 wt .-% SiO ₂ and 40 to 70 wt .-% CaO. 4. The method according to claim 2, characterized in that the mixing or coating of the waste is carried out with a cement containing 25 to 35 wt .-% SiO ₂ , 1 ₀ to 25 wt .-% CaO and 30 to 50 wt .-% % Al ₂ O ₃ contains. 5. The method according to claim 1 or 2, characterized in that the heat treatment of the shaped body (step c) is carried out at a temperature between 150 ° and 250 ° C and over a period of between 12 hours and 4 days.