GB2367419A

GB2367419A - Encapsulation of waste

Info

Publication number: GB2367419A
Application number: GB0020405A
Authority: GB
Inventors: Ewan Robert Maddrell
Original assignee: British Nuclear Fuels PLC
Current assignee: Sellafield Ltd
Priority date: 2000-08-19
Filing date: 2000-08-19
Publication date: 2002-04-03
Also published as: GB0020405D0

Abstract

A means for immobilising both highly active waste and medium active waste arising from an advanced Purex reprocessing plant in which substantial amounts of the non-fuel components of a fuel assembly are also taken into solution during the head-end dissolution step comprises a sodium phosphate glass matrix with zirconia particles distributed within the glass matrix, wherein there is dissolved in the glass at least elements from dissolved nuclear fuel assemblies and cladding, fission products and other radioactive species from irradiated nuclear fuel.

Description

ENCAPSULATION OF WASTE The present invention relates to an immobilising medium for the encapsulation of radioactive waste resulting from the reprocessing of irradiated nuclear fuel and a method for preparing the same.

Nuclear reprocessing plants use the well-established Purex process. These plants produce both highly active (HA) wastes and medium active (MA) wastes. The term HA waste is used generally to mean the bulk of the fission products with associated material from irradiated nuclear fuel. The term MA waste is used generally to mean materials, e. g. fuel cladding, which have gained a substantial measure of radioactivity by contact with the fuel proper or by exposure to the neutron flux but do not generate significant heat levels. The Purex process involves stripping the cladding from the fuel rod or leaching the fuel from inside and then dissolving the spent fuel. The uranium and plutonium from the spent fuel is then separated from the minor actinides and fission products by solvent extraction. The HA wastes comprise the minor actinides and fission products separated from the spent fuel.

MA wastes typically comprise the remains of the stripped or leached nuclear fuel cladding as well as contaminated filters and other contaminated components of fuel assemblies.

In addition, phosphate containing effluents from the operations using tributylphosphate (tbp) solvent during nuclear fuel reprocessing and sodium containing effluents which may arise from the use of solvent cleaning chemicals such as NaOH in the reprocessing plant may be classed as MA waste.

Vitrification has been the preferred method of encapsulating HA wastes. The method involves the incorporation of the waste within a continuous amorphous matrix. Encapsulation in cement has to date been the preferred method of encapsulating MA wastes.

However, waste streams which are likely to arise in the future due to developments to the so-called Purex process (so-called Advanced Purex process) may not be suitable for containment by the vitrification technique due principally to relatively high levels of iron, chromium and zirconium which result from the non-fuel components of fuel assemblies which are also taken into solution in the envisaged new reprocessing techniques.

An envisaged Advanced Purex process employs a different headend process to conventional Purex, i. e. the process of separating the cladding from the fuel and converting the fuel into a form suitable for chemical separation. The general scheme of an envisaged Advanced Purex head-end process and its waste arisings is shown schematically in Figure 1. The whole fuel assembly including stainless steel structural components, cladding and fuel rods is subjected to a nitric acid dissolution step 1 as opposed to just the fuel in conventional Purex. This results in large amounts of iron, nickel, chromium and zirconium in the HA solution in addition to the usual actinides and fission products (FP). About 15% of the Zircaloy (trade mark) fuel cladding and most of the stainless steel and Inconel (trade mark) components are taken into solution along with the irradiated fuel. The remainder of the Zircaloy is left as a slightly oxygen deficient MA zirconia sludge. This sludge is separated and treated as MA waste 4.

The HA solution is then subjected to solvent extraction 3 to separate the uranium and plutonium from the waste actinides, fission products, iron, nickel, chromium and zirconium which are routed to the HA waste stream 5.

An advanced Purex reprocessing plant will also produce MA liquid wastes containing significant quantities of sodium and phosphate. MA waste streams containing sodium, e. g. as sodium nitrate, may arise as a result of using solvent cleaning chemicals such as NaOH in the reprocessing plant. MA waste streams containing phosphate may arise from operations using tributylphosphate (tbp) solvent during nuclear fuel reprocessing. These MA waste streams thus form part of the overall MA waste from the plant 4.

In conventional oxide fuel Purex reprocessing the HA waste consists predominantly of fission products and is immobilised within a vitrified matrix at a waste loading of 20-25 wt%.

The high level waste produced by more modern and improved Advanced Purex reprocessing routes, however, contains such high quantities of inert material from the fuel assembly that vitrification at the same waste loading would roughly quadruple the volume of HA waste produced per tonne of fuel reprocessed.

It is therefore desirable to be able to accommodate higher loadings of active waste into the immobilising medium so as to minimise the volume of the final immobilised waste.

Whilst technologies exist to treat the HA and MA streams separately, it may be convenient to immobilise both the HA wastes and MA wastes arising from an Advanced Purex reprocessing plant in a composite waste form to which all waste and effluent streams could be routed. This has the advantages that the plant only produces a single form of waste and thus two different types of encapsulation lines are not required.

According to a first aspect of the present invention there is provided a waste immobilising medium in which highly radioactive waste is contained, the waste immobilising medium comprising a sodium phosphate glass matrix with zirconia particles distributed within the glass matrix, wherein there is dissolved in the glass at least elements from dissolved nuclear fuel assemblies and cladding, fission products and other radioactive species from irradiated nuclear fuel.

Minor amounts of sodium zirconium phosphate may also be present in the immobilising medium.

The waste immobilising medium is for containing combined HA and MA waste from an advanced Purex reprocessing plant.

The waste immobilising medium is highly durable and leach resistant and is suitable for long term storage of radioactive waste.

The waste immobilising medium enables a waste loading of up to about 70 weight % to be achieved.

The elements from dissolved nuclear fuel assemblies and cladding typically comprise iron, nickel, chromium and zirconium.

The other radioactive species from irradiated nuclear fuel may comprise actinide elements.

The glass matrix efficiently acts as an host for highly radioactive elements, for example the fission products and actinide elements. For example, caesium, barium and strontium may be dissolved in the glass.

The zirconia is not specifically required to act as a host phase.

Preferably, the composition of the sodium phosphate glass has a Na/P molar ratio of between about 1 and 1.5. More preferably, the Na/P molar ratio is around 1.22 (=55/45).

The durability and leach resistance of the glass is enhanced by the presence of dissolved iron, nickel, chromium oxides in the glass which may be dissolved up to their solubility limits. Surplus iron, nickel, chromium and zirconium oxides may exist as discrete particles of spinel and baddeleyite distributed within the glass.

The durability of the glass may also be increased by dissolution of some zirconia within the glass matrix.

Additionally, other components such as alumina, A1203, may be included in the immobilising medium to impart further durability to the glass if required. Up to 20 mol% of the glass may be made up of A1203 in this way.

The waste is a combination of HA waste and MA waste streams arising from an Advanced Purex reprocessing plant utilising a head-end process in which significant amounts of material from e. g. the non-fuel components of the fuel assemblies is dissolved along with the fuel.

The fission products and other radioactive species from irradiated nuclear fuel are predominantly derived from the HA waste material which arises from the solvent extraction cycle in reprocessing. The MA waste may contribute a very minor amount of the fission products and other radioactive species.

The bulk of the zirconia is derived from the waste itself.

Zirconia is typically present in HA and MA wastes from Advanced Purex reprocessing in high amounts from Zircaloy (trade name) fuel cladding as described above.

At least a portion of the sodium and phosphate used to form the sodium phosphate glass comes from MA waste containing sodium and phosphate. The term phosphate is used herein to refer to oxide species of phosphorus generally and not specifically any particular stoichiometry.

Thus, overall, HA wastes containing fission products, iron, nickel, chromium and zirconium and MA wastes containing zirconium, phosphate and sodium are combined and encapsulated in one immobilising medium by the present invention. The zirconium, phosphate and sodium in the wastes being used to form in part the immobilising medium.

The waste immobilising medium may achieve a waste loading of up to about 70 weight % waste. Waste loading is defined as the mass of waste/total mass of waste immobilising medium, which is the same as mass of waste/ (mass of waste + mass of additives).

Such a high waste loading is possible because of the use of sodium, phosphate, zirconia, iron, nickel, and chromium from the waste to form the main phases of the medium. Maximising the waste loading and thereby minimising the final volume of the waste form is one of the key aims of any new waste form.

The iron, nickel and chromium give the glass the necessary durability.

The volume of the final immobilised waste form according to the present invention is about 0.2 m3 per tonne of reprocessed fuel. In addition, all of the HA and MA waste from the plant is held in one immobilising medium. Combining all the waste in one medium means that the overall process for waste encapsulation is simpler as two separate encapsulation methods for HA and MA waste are not required.

According to a second aspect of the present invention there is provided a method of preparing a waste immobilising medium according to the first aspect of the invention, the method including the steps of forming a mixture comprising highly radioactive material, zirconia, phosphate and a sodium containing component ; drying the mixture; calcining the dried mixture; and pressing and sintering the calcined mixture.

Preferably, the amounts of phosphate and sodium are adjusted so that a sodium phosphate glass is formed in the final waste immobilising medium having a Na/P molar ratio of between about 1 and 1.5. More preferably, the glass has a Na/P molar ratio around 1.22 (=55/45).

The highly radioactive material results from the dissolution of fuel assemblies, cladding and fuel in an Advanced Purex reprocessing scheme as described above.

The highly radioactive material is a combination of HA and MA waste streams from an Advanced Purex reprocessing plant.

The highly radioactive material comprises minor actinides and fission products separated from irradiated nuclear fuel by the reprocessing.

The highly radioactive minor actinides and fission products substantially arise from the waste stream of the solvent extraction cycle of reprocessing which separates out the uranium and plutonium.

A minor amount of the minor actinides and fission products may come from the residual radioactive species present in a MA waste which is also immobilised in the medium.

The highly radioactive material is typically provided in the form of a waste liquor. The waste liquor comprises a combination of HA and MA waste streams from an Advanced Purex reprocessing plant.

The waste liquor preferably contains phosphate and a sodiumcontaining component. Thus, the waste liquor may provide at least some of the phosphate and sodium for forming the sodium phosphate glass matrix constituent of the waste immobilising medium. The waste liquor may provide all of at least one of the phosphate and sodium constituents.

The waste liquor typically contains the MA zirconia sludge from the MA waste stream. The waste liquor also contains zirconia from the HA waste stream resulting from dissolution of fuel cladding. The zirconia present in the waste immobilising medium thus predominantly results from this zirconia in the waste.

Supplementary amounts of sodium and phosphate may be added to the waste liquor. This is so that the amounts of the sodium and phosphate are adjusted to enable a glass matrix phase to be formed in the final immobilising medium having the preferred Na/P molar ratio of between about 1 and 1.5 or the more preferred Na/P molar ratio of around 1.22 (=55/45).

The waste liquor also typically comprises substantial amounts of iron, nickel and chromium resulting from dissolution of the fuel assembly and cladding in the Advanced Purex process.

As the waste itself comprises zirconia, phosphate and sodium, the proportions of waste to the supplementary amounts of phosphate and sodium are such that a waste loading of typically 70 weight% may be achieved in the final immobilising medium.

Other components may be added in the mixture. For example, alumina (Al203) may be added. The amount of alumina may be added in an amount equivalent to a concentration of Al203 in the glass of up to 20 mol%. The purpose of the alumina is to add further durability to the glass if required.

The sodium and phosphate may be added as sodium oxide (NazO) or ? 205. More typically, ammonium phosphate or sodium phosphate may be used.

In addition to fission product and actinide elements, the waste liquor may contain gadolinium from its use as a neutron poison in the fuel.

In the waste liquor many of the highly active waste elements may be present in the form of nitrates because of the use of nitric acid in the reprocessing operations.

Preferably, the waste liquor is denitrated before or whilst forming the mixture. This makes further processing of the waste liquor easier. If the liquor is not denitrated, an undesirable sludge or paste may be formed in the mixture which may be difficult to dry effectively.

The denitration may be performed in one of many ways. A preferred method of denitration comprises reacting the liquor with formaldehyde. After denitration, the liquor remains as a substantially liquid phase.

Mixing of the components in the mixture is effected typically by stirring. Stirring ensures homogeneity in the mixture.

Other methods of homogeneously mixing may be used.

After the mixture has been formed and sufficiently mixed, the mixture is dried. The drying may be carried out by one of many methods known to the person skilled in the art. For example, a hot-plate or similar could be used.

After the mixture has been dried, it is calcined to form a powder. The calcination may be carried out in a neutral (e. g. with N2 gas) or reducing atmosphere. The reducing atmosphere may comprise an Ar/H2 mixture or a N2/H2 mixture. The hydrogen is typically diluted to 10% or less in the inert gas. For example, a 5% mixture of H2 in N2 may be used.

The calcination may be carried out between 650-800oC.

Typically, about 750oC may be used.

Optionally, the calcined powder, particularly powder calcined in an N2/H2 mixture, may be mixed with an oxygen getter prior to compaction and sintering. The oxygen getter may be a metal. For example, metallic titanium is an effective getter.

Where a metal getter is used, e. g. titanium, it may be present in the powder in an amount of, for example, about 2 wt%.

Finally, the calcined powder is compacted and sintered to produce the final immobilising medium suitable for long term storage.

The compaction and sintering may be carried out according to known methods such as Hot Uniaxial Pressing or Hot Isostatic Pressing (HIP). HIP is preferred. When HIP is used, because the sintering is under pressure and there is no volatilisation problem, higher temperatures may be used which enables more Fe, Cr, Ni, Zr to be taken into the glass making it more durable. Thus, referably the temperature for HIP is 1000 1400oC. More preferably the temperature for HIP is 1100 1300oC.

Specific embodiments of the present invention will now be described by way of example and with reference to Figure 1 which shows an X-Ray Diffraction (XRD) pattern of a sample of a waste immobilising medium according to the present invention. The embodiments are illustrative only and do not limit the invention in any way.

Example 1 The compositions of various envisaged wastes are given below in Table 1. They simulate the waste arisings for one tonne of nuclear fuel being reprocessed in an advanced Purex reprocessing plant and contain substantial amounts of zirconia, phosphate and sodium.

TABLE I WASTE AND EFFLUENT ARISINGS PER TONNE FUEL*

HA WASTE OXIDES HBU MOX 3: 1 mix HBU: MOX Fission 58. 2 57. 7 58. 1 Products Gadolinium 9. 1 9. 1 9. 1 Fe203 63. 8 63. 8 63. 8 Cr20321. 021. 021. 0 NiO 19. 5 19. 5 19. 5 ZrO2 49. 9 49. 9 49. 9 MOO3 0. 8 0. 8 0. 8 SnO2 0.8 0.8 0. 8 Actinides 2. 2 6. 6 3. 3 MA WASTE OXIDES ZrO2 as 287. 5 287. 5 287. 5 Zirconia sludge Phosphate as 20 20 20 P205 Sodium as Na20 35 35 35 HBU = High Burn Up uranium dioxide fuel MOX = mixed oxide fuel * all values are in kg.

A waste immobilising medium was prepared as follows using a simulated waste from reprocessing 3: 1 HBU: MOX mixed fuel as listed in the last column of Table 1 having a burn-up of 55GWd/te (GWd/te = giga watt days per tonne).

The oxides of the waste elements were then obtained by denitrating solutions of the corresponding nitrates as described above.

Additional NazO and P205 were added to the waste solution after the denitration but before drying (to ensure homogeneous mixing) in the following amounts per one tonne fuel reprocessed.

Additives per tonne fuel reprocessed : Na20 66.3 kg PzOs 169.9 kg Batches were prepared corresponding to 1/2000th of the waste arisings per tonne.

After drying, the dried mixture was divided into two portions before calcination. The two portions were then each calcined at 750oC for 4 hours. One portion was calcined in flowing nitrogen, the other in flowing nitrogen/5% hydrogen. The gas flow rate was about 1 litre/minute.

The calcined mixtures were then ball milled to break down coarse aggregates and then they were hot isostatically pressed at 1200oC and a pressure of 200 Mpa for 4 hours.

The pressed waste forms were then decanned and characterised by X-ray diffraction (XRD).

An XRD pattern is shown in Figure 2. Figure 2 also shows the theoretical peak positions calculated for given phases as indicated by reference numerals 1 and 2. The theoretical peaks fit well with the experimental data. The data shows the presence of at least baddelyite (zero2) and sodium zirconium phosphate (NaZr2 (P04) 3).

Claims

Claims 1. A waste immobilising medium in which highly radioactive waste is contained, the waste immobilising medium comprising a sodium phosphate glass matrix with zirconia particles distributed within the glass matrix, wherein there is dissolved in the glass at least elements from dissolved nuclear fuel assemblies and cladding, fission products and other radioactive species from irradiated nuclear fuel.
2. A waste immobilising medium as in claim 1 wherein the composition of the sodium phosphate glass has a Na/P molar ratio of between about 1 and 1.5.
3. A waste immobilising medium as in claim 2 wherein the Na/P molar ratio is around 1.22.
4. A waste immobilising medium as in any one of claims 1 to 3 wherein the elements from dissolved nuclear fuel assemblies and cladding dissolved in the glass matrix comprise iron, nickel and chromium.
5. A waste immobilising medium as in any one of claims 1 to 4 wherein the other radioactive species from irradiated nuclear fuel comprise actinide elements.
6. A waste immobilising medium as in any one of claims 1 to 5 wherein the fission products and other radioactive species from irradiated nuclear fuel are predominantly derived from a highly radioactive waste material produced by the solvent extraction cycle in reprocessing.
7. A waste immobilising medium as in any one of claims 1 to 6 wherein a minor proportion of the fission products and other radioactive species from irradiated nuclear fuel are derived from a medium active waste.
8. A waste immobilising medium as in any one of claims 1 to 7 wherein the medium further comprises a minor amount of sodium zirconium phosphate.
9. A waste immobilising medium as in any one of claims 1 to 8 wherein aluminium oxide is dissolved in the glass matrix.
10. A waste immobilising medium as in any one of claims 1 to 9 wherein at least some of the zirconia or sodium or phosphate of the sodium phosphate glass originates from the radioactive waste.
11. A waste immobilising medium as in any one of claims 1 to 10 wherein the waste loading is about 70 weight % waste or less.
12. A waste immobilising medium as in claim 1 and substantially as herein described.
13. A waste immobilising medium substantially as herein described with reference to the examples.
14. A method of preparing a waste immobilising medium as in claims 1 to 13, the method including the steps of forming a mixture comprising highly radioactive material, zirconia, phosphate and sodium; drying the mixture; calcining the dried mixture ; and pressing and sintering the calcined mixture.
15. A method of preparing a waste immobilising medium as in claim 14 wherein a radioactive waste liquor provides the highly radioactive material and a substantial amount of the zirconia, phosphate and sodium.
16. A method of preparing a waste immobilising medium as in claim 14 or 15 wherein the highly radioactive material is predominantly derived from a highly radioactive waste material produced by a solvent extraction cycle in reprocessing.
17. A method of preparing a waste immobilising medium as in claim 16 wherein the highly radioactive material also comprises a minor amount of highly radioactive material from a medium active waste.
18. A method of preparing a waste immobilising medium as in any one of claims 14 to 17 wherein the radioactive waste is eventually contained in the waste immobilising medium at a 70 weight% loading or less.
19. A method of preparing a waste immobilising medium substantially as herein described with reference to the examples.
20. A method of preparing a waste immobilising medium as in claim 12 and substantially as herein described.