GB2367420A - Encapsulation of waste - Google Patents
Encapsulation of waste Download PDFInfo
- Publication number
- GB2367420A GB2367420A GB0020406A GB0020406A GB2367420A GB 2367420 A GB2367420 A GB 2367420A GB 0020406 A GB0020406 A GB 0020406A GB 0020406 A GB0020406 A GB 0020406A GB 2367420 A GB2367420 A GB 2367420A
- Authority
- GB
- United Kingdom
- Prior art keywords
- waste
- medium
- sodium
- phosphate
- immobilising medium
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Withdrawn
Links
- 239000002699 waste material Substances 0.000 title claims abstract description 90
- 238000005538 encapsulation Methods 0.000 title description 4
- MCMNRKCIXSYSNV-UHFFFAOYSA-N Zirconium dioxide Chemical compound O=[Zr]=O MCMNRKCIXSYSNV-UHFFFAOYSA-N 0.000 claims abstract description 50
- 239000011521 glass Substances 0.000 claims abstract description 19
- 239000011159 matrix material Substances 0.000 claims abstract description 17
- 239000002245 particle Substances 0.000 claims abstract description 3
- 239000001488 sodium phosphate Substances 0.000 claims abstract description 3
- 229910000162 sodium phosphate Inorganic materials 0.000 claims abstract description 3
- RYFMWSXOAZQYPI-UHFFFAOYSA-K trisodium phosphate Chemical compound [Na+].[Na+].[Na+].[O-]P([O-])([O-])=O RYFMWSXOAZQYPI-UHFFFAOYSA-K 0.000 claims abstract description 3
- 239000011734 sodium Substances 0.000 claims description 33
- 229910019142 PO4 Inorganic materials 0.000 claims description 28
- NBIIXXVUZAFLBC-UHFFFAOYSA-K phosphate Chemical compound [O-]P([O-])([O-])=O NBIIXXVUZAFLBC-UHFFFAOYSA-K 0.000 claims description 28
- 239000010452 phosphate Substances 0.000 claims description 28
- 239000000203 mixture Substances 0.000 claims description 27
- DGAQECJNVWCQMB-PUAWFVPOSA-M Ilexoside XXIX Chemical compound C[C@@H]1CC[C@@]2(CC[C@@]3(C(=CC[C@H]4[C@]3(CC[C@@H]5[C@@]4(CC[C@@H](C5(C)C)OS(=O)(=O)[O-])C)C)[C@@H]2[C@]1(C)O)C)C(=O)O[C@H]6[C@@H]([C@H]([C@@H]([C@H](O6)CO)O)O)O.[Na+] DGAQECJNVWCQMB-PUAWFVPOSA-M 0.000 claims description 26
- 229910052708 sodium Inorganic materials 0.000 claims description 26
- 238000000034 method Methods 0.000 claims description 24
- PNEYBMLMFCGWSK-UHFFFAOYSA-N Alumina Chemical compound [O-2].[O-2].[O-2].[Al+3].[Al+3] PNEYBMLMFCGWSK-UHFFFAOYSA-N 0.000 claims description 11
- 239000002901 radioactive waste Substances 0.000 claims description 9
- AJPJDKMHJJGVTQ-UHFFFAOYSA-M sodium dihydrogen phosphate Chemical compound [Na+].OP(O)([O-])=O AJPJDKMHJJGVTQ-UHFFFAOYSA-M 0.000 claims description 9
- 238000011068 loading method Methods 0.000 claims description 8
- 238000001354 calcination Methods 0.000 claims description 5
- 238000001035 drying Methods 0.000 claims description 5
- 238000005245 sintering Methods 0.000 claims description 5
- 239000012857 radioactive material Substances 0.000 claims description 4
- 238000003825 pressing Methods 0.000 claims description 3
- YHKRPJOUGGFYNB-UHFFFAOYSA-K sodium;zirconium(4+);phosphate Chemical compound [Na+].[Zr+4].[O-]P([O-])([O-])=O YHKRPJOUGGFYNB-UHFFFAOYSA-K 0.000 claims description 3
- 239000011363 dried mixture Substances 0.000 claims description 2
- 230000002285 radioactive effect Effects 0.000 claims description 2
- 239000000446 fuel Substances 0.000 abstract description 21
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 abstract description 17
- 238000012958 reprocessing Methods 0.000 abstract description 15
- 238000004090 dissolution Methods 0.000 abstract description 4
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 8
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 description 8
- 238000005253 cladding Methods 0.000 description 7
- 230000004992 fission Effects 0.000 description 7
- 239000000047 product Substances 0.000 description 7
- HEMHJVSKTPXQMS-UHFFFAOYSA-M Sodium hydroxide Chemical compound [OH-].[Na+] HEMHJVSKTPXQMS-UHFFFAOYSA-M 0.000 description 6
- 229910052768 actinide Inorganic materials 0.000 description 6
- 150000001255 actinides Chemical class 0.000 description 6
- 239000003758 nuclear fuel Substances 0.000 description 6
- 239000000843 powder Substances 0.000 description 5
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 description 4
- KKCBUQHMOMHUOY-UHFFFAOYSA-N Na2O Inorganic materials [O-2].[Na+].[Na+] KKCBUQHMOMHUOY-UHFFFAOYSA-N 0.000 description 4
- 229910052804 chromium Inorganic materials 0.000 description 4
- 239000011651 chromium Substances 0.000 description 4
- 238000001513 hot isostatic pressing Methods 0.000 description 4
- 229910052742 iron Inorganic materials 0.000 description 4
- 229910052759 nickel Inorganic materials 0.000 description 4
- 239000002904 solvent Substances 0.000 description 4
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 3
- 229910001093 Zr alloy Inorganic materials 0.000 description 3
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 3
- 239000000463 material Substances 0.000 description 3
- 238000002156 mixing Methods 0.000 description 3
- 239000001301 oxygen Substances 0.000 description 3
- 229910052760 oxygen Inorganic materials 0.000 description 3
- 239000010802 sludge Substances 0.000 description 3
- 239000002915 spent fuel radioactive waste Substances 0.000 description 3
- 239000000126 substance Substances 0.000 description 3
- 229910052726 zirconium Inorganic materials 0.000 description 3
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 description 2
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 2
- 229910052778 Plutonium Inorganic materials 0.000 description 2
- RTAQQCXQSZGOHL-UHFFFAOYSA-N Titanium Chemical compound [Ti] RTAQQCXQSZGOHL-UHFFFAOYSA-N 0.000 description 2
- 229910052770 Uranium Inorganic materials 0.000 description 2
- 238000002441 X-ray diffraction Methods 0.000 description 2
- 239000000654 additive Substances 0.000 description 2
- 230000000712 assembly Effects 0.000 description 2
- 238000000429 assembly Methods 0.000 description 2
- 238000004140 cleaning Methods 0.000 description 2
- 238000005056 compaction Methods 0.000 description 2
- 239000000470 constituent Substances 0.000 description 2
- 229910052593 corundum Inorganic materials 0.000 description 2
- 239000007789 gas Substances 0.000 description 2
- 239000001257 hydrogen Substances 0.000 description 2
- 229910052739 hydrogen Inorganic materials 0.000 description 2
- 239000010808 liquid waste Substances 0.000 description 2
- 230000007774 longterm Effects 0.000 description 2
- 229910052751 metal Inorganic materials 0.000 description 2
- 239000002184 metal Substances 0.000 description 2
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 2
- VWDWKYIASSYTQR-UHFFFAOYSA-N sodium nitrate Chemical compound [Na+].[O-][N+]([O-])=O VWDWKYIASSYTQR-UHFFFAOYSA-N 0.000 description 2
- 238000000638 solvent extraction Methods 0.000 description 2
- 229910001220 stainless steel Inorganic materials 0.000 description 2
- 239000010935 stainless steel Substances 0.000 description 2
- 238000003756 stirring Methods 0.000 description 2
- 238000003860 storage Methods 0.000 description 2
- 239000010936 titanium Substances 0.000 description 2
- 229910052719 titanium Inorganic materials 0.000 description 2
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 2
- 229940093635 tributyl phosphate Drugs 0.000 description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 2
- 229910001845 yogo sapphire Inorganic materials 0.000 description 2
- 239000004254 Ammonium phosphate Substances 0.000 description 1
- 229910052688 Gadolinium Inorganic materials 0.000 description 1
- 229910002651 NO3 Inorganic materials 0.000 description 1
- NHNBFGGVMKEFGY-UHFFFAOYSA-N Nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 description 1
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 1
- 230000000996 additive effect Effects 0.000 description 1
- 229910000148 ammonium phosphate Inorganic materials 0.000 description 1
- 235000019289 ammonium phosphates Nutrition 0.000 description 1
- 239000004568 cement Substances 0.000 description 1
- 230000002950 deficient Effects 0.000 description 1
- 238000011161 development Methods 0.000 description 1
- 230000018109 developmental process Effects 0.000 description 1
- MNNHAPBLZZVQHP-UHFFFAOYSA-N diammonium hydrogen phosphate Chemical compound [NH4+].[NH4+].OP([O-])([O-])=O MNNHAPBLZZVQHP-UHFFFAOYSA-N 0.000 description 1
- 230000004907 flux Effects 0.000 description 1
- UIWYJDYFSGRHKR-UHFFFAOYSA-N gadolinium atom Chemical compound [Gd] UIWYJDYFSGRHKR-UHFFFAOYSA-N 0.000 description 1
- 229910001026 inconel Inorganic materials 0.000 description 1
- 238000010348 incorporation Methods 0.000 description 1
- 239000011261 inert gas Substances 0.000 description 1
- 238000002386 leaching Methods 0.000 description 1
- 238000004519 manufacturing process Methods 0.000 description 1
- 230000007935 neutral effect Effects 0.000 description 1
- 229910017604 nitric acid Inorganic materials 0.000 description 1
- 229910052757 nitrogen Inorganic materials 0.000 description 1
- 238000009376 nuclear reprocessing Methods 0.000 description 1
- 239000002574 poison Substances 0.000 description 1
- 231100000614 poison Toxicity 0.000 description 1
- 238000002360 preparation method Methods 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 238000000926 separation method Methods 0.000 description 1
- 235000010344 sodium nitrate Nutrition 0.000 description 1
- 239000004317 sodium nitrate Substances 0.000 description 1
- 229910001948 sodium oxide Inorganic materials 0.000 description 1
- 239000013589 supplement Substances 0.000 description 1
- 238000003826 uniaxial pressing Methods 0.000 description 1
- 238000004017 vitrification Methods 0.000 description 1
Classifications
-
- C—CHEMISTRY; METALLURGY
- C03—GLASS; MINERAL OR SLAG WOOL
- C03C—CHEMICAL COMPOSITION OF GLASSES, GLAZES OR VITREOUS ENAMELS; SURFACE TREATMENT OF GLASS; SURFACE TREATMENT OF FIBRES OR FILAMENTS MADE FROM GLASS, MINERALS OR SLAGS; JOINING GLASS TO GLASS OR OTHER MATERIALS
- C03C1/00—Ingredients generally applicable to manufacture of glasses, glazes, or vitreous enamels
- C03C1/002—Use of waste materials, e.g. slags
-
- C—CHEMISTRY; METALLURGY
- C03—GLASS; MINERAL OR SLAG WOOL
- C03C—CHEMICAL COMPOSITION OF GLASSES, GLAZES OR VITREOUS ENAMELS; SURFACE TREATMENT OF GLASS; SURFACE TREATMENT OF FIBRES OR FILAMENTS MADE FROM GLASS, MINERALS OR SLAGS; JOINING GLASS TO GLASS OR OTHER MATERIALS
- C03C14/00—Glass compositions containing a non-glass component, e.g. compositions containing fibres, filaments, whiskers, platelets, or the like, dispersed in a glass matrix
- C03C14/004—Glass compositions containing a non-glass component, e.g. compositions containing fibres, filaments, whiskers, platelets, or the like, dispersed in a glass matrix the non-glass component being in the form of particles or flakes
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/28—Treating solids
- G21F9/30—Processing
- G21F9/301—Processing by fixation in stable solid media
- G21F9/302—Processing by fixation in stable solid media in an inorganic matrix
- G21F9/305—Glass or glass like matrix
-
- C—CHEMISTRY; METALLURGY
- C03—GLASS; MINERAL OR SLAG WOOL
- C03C—CHEMICAL COMPOSITION OF GLASSES, GLAZES OR VITREOUS ENAMELS; SURFACE TREATMENT OF GLASS; SURFACE TREATMENT OF FIBRES OR FILAMENTS MADE FROM GLASS, MINERALS OR SLAGS; JOINING GLASS TO GLASS OR OTHER MATERIALS
- C03C2214/00—Nature of the non-vitreous component
- C03C2214/14—Waste material, e.g. to be disposed of
Landscapes
- Chemical & Material Sciences (AREA)
- Engineering & Computer Science (AREA)
- Chemical Kinetics & Catalysis (AREA)
- Life Sciences & Earth Sciences (AREA)
- General Chemical & Material Sciences (AREA)
- Geochemistry & Mineralogy (AREA)
- Materials Engineering (AREA)
- Organic Chemistry (AREA)
- Physics & Mathematics (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Ceramic Engineering (AREA)
- Dispersion Chemistry (AREA)
- Inorganic Chemistry (AREA)
- Processing Of Solid Wastes (AREA)
Abstract
A means for immobilising 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 based glass matrix with zirconia particles distributed therein.
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 differ greatly in their composition and physical nature due principally to relatively high levels of iron, chromium, nickel 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 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 4 thus form part of the overall MA waste from the plant.
With the advent of the new form of MA waste soon to arise from advanced Purex reprocessing operations there exists a need to improve the final waste form to combine the characteristics of durability, leach resistance over long periods of time, high loading of radioactive waste and ease and cost effectiveness of manufacture.
The present invention aims to provide a suitable manner of immobilising MA waste arising from advanced Purex reprocessing operations.
According to a first aspect of the present invention there is provided a waste immobilising medium in which a medium radioactive waste is contained, the waste immobilising medium comprising a sodium phosphate based glass matrix with zirconia particles distributed therein.
The radioactive waste is a MA waste containing substantial quantities of zirconia arising from advanced Purex reprocessing operations.
Minor amounts of a crystalline sodium zirconium phosphate phase may also be present within the immobilising medium.
The waste immobilising medium is highly durable and leach resistant. Thus, the immobilising medium is suitable for long term storage of radioactive waste.
The glass matrix may efficiently act as a host for any small amounts of radioactive elements, for example fission product and actinide elements, which may be present in the MA waste.
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 of the glass may be increased by dissolution of some zirconia within the glass matrix.
Additionally, alumina, Al203, may be included in the immobilising medium to impart durability to the glass. In such cases the alumina is typically incorporated in the glass matrix as part of a sodium aluminophosphate glass phase.
Where A1203 is added to the glass to impart additional durability, the A1203 may be present in an amount up to 20 mol% of the glass matrix.
If any small amounts of iron, chromium and nickel are present in the MA waste from the advanced Purex dissolution step, these may be dissolved in the glass. These may also impart durability to the glass.
The zirconia in the immobilising medium originates from the MA waste itself as it comes from the remains of dissolved
Zircaloy (trade name) fuel cladding.
As an advanced Purex reprocessing plant may also produce medium active liquid wastes containing significant quantities of sodium and phosphate these components may also be utilised by the present invention. Advantageously, the waste may provide at least some of the sodium and phosphate used to form the sodium phosphate glass.
Thus MA waste streams from an advanced Purex reprocessing plant containing zirconia, sodium and phosphate may be combined and encapsulated together in the immobilising medium.
The waste itself in such a case provides much of the material for the constituent sodium phosphate glass matrix and the zirconia.
The waste immobilising medium may achieve a waste loading of about 85 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).
The use of zirconia, sodium and phosphate from the waste to form the phases of the medium enables high waste loadings to be achieved. The higher the waste loading achieved, the smaller the final volume of the waste form will be.
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 zirconia, phosphate, sodium and a medium radioactive material; drying the mixture; calcining the dried mixture; and pressing and sintering the calcined mixture.
Preferably, the amounts of zirconia, 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 Na/P molar ratio is around 1.22 (=55/45).
The radioactive material is typically provided in the form of a waste liquor.
The waste liquor typically contains substantial amounts of zirconia in the form of a sludge. The waste liquor preferably contains phosphate and sodium. Thus, the waste liquor may provide some of the zirconia, phosphate and sodium for forming the zirconia and sodium phosphate glass matrix constituents of the waste immobilising medium. The waste liquor may provide all of at least one of the zirconia, phosphate and sodium.
Supplementary amounts of sodium and phosphate are typically 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 ratios described above.
Where 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 up to 85 weight% may be achieved in the final immobilising medium.
Other components may be added in the mixture. Preferably, alumina, Ail203, is added. The alumina imparts durability to the glass matrix in the final immobilising medium. The amount of alumina may be added in an amount equivalent to a
concentration of Ail203 in the glass of up to 20 mol%. A typical glass matrix composition may be 44 mol% Na20, 36 mol% P205 and 20 mol% Al203.
The sodium may be added as sodium oxide, Na20. The phosphate may be added as P2Os or more typically as ammonium phosphate.
The waste liquor may contain small amounts of fission product elements. Gadolinium may also be present in small amounts in the waste liquor from its use as a neutron poison in the fuel.
There may be traces of actinides present.
If required, the waste liquor may be denitrated in a known way before further processing. However, the MA waste stream from reprocessing is typically not high enough in nitrate to necessitate denitration.
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 skilled person in the art. For example, the drying may be effected on a hot plate on similar apparatus.
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. However, a hydrogen atmosphere may be unnecessary and an atmosphere such as air could be used for calcining.
The calcination may be carried out between 650-800 C.
Typically, about 750 C 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. Preferably the temperature for HIP is 1000-14000C. More preferably the temperature for
HIP is 1100-1300 C.
Specific embodiments of the present invention will now be described by way of example. The embodiments are illustrative only and do not limit the invention in any way.
Example 1
Waste immobilising media were prepared as follows using a simulated waste. The composition of the simulated waste used is given below in Table 1. It simulates the medium-active waste arisings for one tonne of nuclear fuel being reprocessed in an advanced Purex reprocessing plant and contains substantial amounts of zirconia, phosphate and sodium.
Table 1: simulated waste composition
Waste Component Waste amount (Kg) arising per tonne of reprocessed fuel Zirconia, 285 Phosphate, P2O5 20 Sodium, as Na20 35 Further phosphate, P205, was also added in an amount equal to 55-60 Kg per tonne of fuel reprocessed to supplement the phosphate in the waste.
The mixture of simulated waste plus phosphate additive was prepared by mixing solutions of zirconia, phosphate and sodium.
Batches of the mixture in amounts equal to 1/2000th of the arisings per tonne of reprocessed fuel were prepared and used in the preparations.
The mixture was then stir-dried. After drying, the mixture was calcined at 750oC for 4 hours in flowing nitrogen (air could also be used). 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 2 hours.
The pressed waste forms were then decanned and characterised by X-ray diffraction (XRD). The XRD analysis showed the presence of baddelyite (ZrO2) and sodium zirconium phosphate (NaZr2 (P04) 3).
Claims (13)
- Claims 1. A waste immobilising medium in which a medium radioactive waste is contained, the waste immobilising medium comprising a sodium phosphate based glass matrix with zirconia particles distributed therein.
- 2. A waste immobilising medium as in claim 1 wherein the sodium phosphate glass matrix has a Na/P molar ratio of between about 1 and 1.5.
- 3. A waste immobilising medium as in claim 2 wherein the sodium phosphate glass matrix has a Na/P molar ratio of around 1.22
- 4. A waste immobilising medium as in any one of claims 1 to 3 wherein minor quantities of radioactive elements are dissolved in the glass matrix.
- 5. A waste immobilising medium as in any one of claims 1 to 4 wherein the medium further comprises some sodium zirconium phosphate.
- 6. A waste immobilising medium as in any one of claims 1 to 5 wherein aluminium oxide is dissolved in the glass matrix.
- 7. A waste immobilising medium as in any one of claims 1 to 6 wherein at least some of the zirconia or sodium or phosphate of the sodium phosphate glass originates from the radioactive waste.
- 8. A waste immobilising medium as in any one of claims 1 to 7 wherein the waste loading is about 85 weight % waste or less.
- 9. A method of preparing a waste immobilising medium as in claim 1, the method including the steps of forming a mixture comprising zirconia, phosphate, sodium and a medium radioactive material; drying the mixture; calcining the dried mixture; and pressing and sintering the calcined mixture.
- 10. A method of preparing a waste immobilising medium as in claim 9 wherein the amounts of phosphate and sodium are adjusted so that a sodium phosphate glass is formed after the pressing and sintering having sodium and phosphate in the proportions in claim 2 or 3.
- 11. A method of preparing a waste immobilising medium as in claim 9 or 10 wherein a radioactive waste liquor provides the radioactive material and a substantial amount of the zirconia, phosphate and sodium.
- 12. A method of preparing a waste immobilising medium as in claim 11 wherein the radioactive waste is eventually contained in the waste immobilising medium at a 85 weight% loading or less.
- 13. A method of preparing a waste immobilising medium as in any of claims 10 to 12 wherein alumina is also present in the mixture.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| GB0020406A GB2367420A (en) | 2000-08-19 | 2000-08-19 | Encapsulation of waste |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| GB0020406A GB2367420A (en) | 2000-08-19 | 2000-08-19 | Encapsulation of waste |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| GB0020406D0 GB0020406D0 (en) | 2000-10-04 |
| GB2367420A true GB2367420A (en) | 2002-04-03 |
Family
ID=9897887
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| GB0020406A Withdrawn GB2367420A (en) | 2000-08-19 | 2000-08-19 | Encapsulation of waste |
Country Status (1)
| Country | Link |
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| GB (1) | GB2367420A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN109650726A (en) * | 2019-02-21 | 2019-04-19 | 西南科技大学 | The one-step preppn process of sodium zirconium phosphate glass ceramics curing substrate |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB2163893A (en) * | 1984-07-31 | 1986-03-05 | Agip Spa | Immobilising the fission product and transuranic element content of liquid high level radioactive waste |
| WO2001035422A2 (en) * | 1999-11-12 | 2001-05-17 | British Nuclear Fuels Plc | Encapsulation of waste |
-
2000
- 2000-08-19 GB GB0020406A patent/GB2367420A/en not_active Withdrawn
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB2163893A (en) * | 1984-07-31 | 1986-03-05 | Agip Spa | Immobilising the fission product and transuranic element content of liquid high level radioactive waste |
| WO2001035422A2 (en) * | 1999-11-12 | 2001-05-17 | British Nuclear Fuels Plc | Encapsulation of waste |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN109650726A (en) * | 2019-02-21 | 2019-04-19 | 西南科技大学 | The one-step preppn process of sodium zirconium phosphate glass ceramics curing substrate |
Also Published As
| Publication number | Publication date |
|---|---|
| GB0020406D0 (en) | 2000-10-04 |
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| WAP | Application withdrawn, taken to be withdrawn or refused ** after publication under section 16(1) |