GB2367419A - Encapsulation of waste - Google Patents
Encapsulation of waste Download PDFInfo
- Publication number
- GB2367419A GB2367419A GB0020405A GB0020405A GB2367419A GB 2367419 A GB2367419 A GB 2367419A GB 0020405 A GB0020405 A GB 0020405A GB 0020405 A GB0020405 A GB 0020405A GB 2367419 A GB2367419 A GB 2367419A
- Authority
- GB
- United Kingdom
- Prior art keywords
- waste
- medium
- immobilising medium
- radioactive
- sodium
- 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- Withdrawn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 135
- 238000005538 encapsulation Methods 0.000 title description 7
- MCMNRKCIXSYSNV-UHFFFAOYSA-N Zirconium dioxide Chemical compound O=[Zr]=O MCMNRKCIXSYSNV-UHFFFAOYSA-N 0.000 claims abstract description 45
- 238000012958 reprocessing Methods 0.000 claims abstract description 25
- 239000011521 glass Substances 0.000 claims abstract description 22
- 230000004992 fission Effects 0.000 claims abstract description 20
- 239000003758 nuclear fuel Substances 0.000 claims abstract description 20
- 238000005253 cladding Methods 0.000 claims abstract description 15
- 239000011159 matrix material Substances 0.000 claims abstract description 14
- 230000002285 radioactive effect Effects 0.000 claims abstract description 12
- 238000000429 assembly Methods 0.000 claims abstract description 9
- 230000000712 assembly Effects 0.000 claims abstract description 9
- AJPJDKMHJJGVTQ-UHFFFAOYSA-M sodium dihydrogen phosphate Chemical compound [Na+].OP(O)([O-])=O AJPJDKMHJJGVTQ-UHFFFAOYSA-M 0.000 claims abstract description 9
- 239000002245 particle Substances 0.000 claims abstract description 4
- 238000000034 method Methods 0.000 claims description 32
- 239000011734 sodium Substances 0.000 claims description 30
- 229910019142 PO4 Inorganic materials 0.000 claims description 24
- NBIIXXVUZAFLBC-UHFFFAOYSA-K phosphate Chemical compound [O-]P([O-])([O-])=O NBIIXXVUZAFLBC-UHFFFAOYSA-K 0.000 claims description 24
- 239000010452 phosphate Substances 0.000 claims description 24
- XEEYBQQBJWHFJM-UHFFFAOYSA-N Iron Chemical compound [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 claims description 23
- 239000000203 mixture Substances 0.000 claims description 23
- 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 22
- 229910052708 sodium Inorganic materials 0.000 claims description 22
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 claims description 21
- 229910052768 actinide Inorganic materials 0.000 claims description 12
- 150000001255 actinides Chemical class 0.000 claims description 12
- 229910052742 iron Inorganic materials 0.000 claims description 12
- 229910052804 chromium Inorganic materials 0.000 claims description 11
- 239000011651 chromium Substances 0.000 claims description 11
- 238000011068 loading method Methods 0.000 claims description 11
- 229910052759 nickel Inorganic materials 0.000 claims description 11
- VYZAMTAEIAYCRO-UHFFFAOYSA-N Chromium Chemical compound [Cr] VYZAMTAEIAYCRO-UHFFFAOYSA-N 0.000 claims description 10
- 239000012857 radioactive material Substances 0.000 claims description 10
- PNEYBMLMFCGWSK-UHFFFAOYSA-N Alumina Chemical compound [O-2].[O-2].[O-2].[Al+3].[Al+3] PNEYBMLMFCGWSK-UHFFFAOYSA-N 0.000 claims description 9
- 239000002901 radioactive waste Substances 0.000 claims description 9
- 239000000463 material Substances 0.000 claims description 6
- 238000000638 solvent extraction Methods 0.000 claims description 6
- 238000001354 calcination Methods 0.000 claims description 5
- 238000001035 drying Methods 0.000 claims description 5
- 238000005245 sintering Methods 0.000 claims description 5
- 239000011363 dried mixture Substances 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
- 238000003825 pressing Methods 0.000 claims description 2
- 239000000446 fuel Substances 0.000 abstract description 35
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 abstract description 21
- 238000004090 dissolution Methods 0.000 abstract description 6
- 239000000047 product Substances 0.000 description 16
- 229910052726 zirconium Inorganic materials 0.000 description 8
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 7
- HEMHJVSKTPXQMS-UHFFFAOYSA-M Sodium hydroxide Chemical compound [OH-].[Na+] HEMHJVSKTPXQMS-UHFFFAOYSA-M 0.000 description 6
- 238000001513 hot isostatic pressing Methods 0.000 description 5
- 239000000843 powder Substances 0.000 description 5
- 239000010802 sludge Substances 0.000 description 5
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 description 4
- 239000012071 phase Substances 0.000 description 4
- 239000002904 solvent Substances 0.000 description 4
- WSFSSNUMVMOOMR-UHFFFAOYSA-N Formaldehyde Chemical compound O=C WSFSSNUMVMOOMR-UHFFFAOYSA-N 0.000 description 3
- 229910052778 Plutonium Inorganic materials 0.000 description 3
- 229910052770 Uranium Inorganic materials 0.000 description 3
- 238000002441 X-ray diffraction Methods 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
- 238000002156 mixing Methods 0.000 description 3
- 239000001301 oxygen Substances 0.000 description 3
- 229910052760 oxygen Inorganic materials 0.000 description 3
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 3
- KKCBUQHMOMHUOY-UHFFFAOYSA-N sodium oxide Chemical compound [O-2].[Na+].[Na+] KKCBUQHMOMHUOY-UHFFFAOYSA-N 0.000 description 3
- 239000002915 spent fuel radioactive waste Substances 0.000 description 3
- 239000000126 substance Substances 0.000 description 3
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 3
- 238000004017 vitrification Methods 0.000 description 3
- 101100348017 Drosophila melanogaster Nazo gene Proteins 0.000 description 2
- 229910052688 Gadolinium Inorganic materials 0.000 description 2
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 2
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 2
- RTAQQCXQSZGOHL-UHFFFAOYSA-N Titanium Chemical compound [Ti] RTAQQCXQSZGOHL-UHFFFAOYSA-N 0.000 description 2
- 239000000654 additive Substances 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
- UIWYJDYFSGRHKR-UHFFFAOYSA-N gadolinium atom Chemical compound [Gd] UIWYJDYFSGRHKR-UHFFFAOYSA-N 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
- 230000007774 longterm Effects 0.000 description 2
- 229910052751 metal Inorganic materials 0.000 description 2
- 239000002184 metal Substances 0.000 description 2
- 150000002823 nitrates Chemical class 0.000 description 2
- 229910017604 nitric acid Inorganic materials 0.000 description 2
- 229910052757 nitrogen Inorganic materials 0.000 description 2
- VWDWKYIASSYTQR-UHFFFAOYSA-N sodium nitrate Chemical compound [Na+].[O-][N+]([O-])=O VWDWKYIASSYTQR-UHFFFAOYSA-N 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
- XOLBLPGZBRYERU-UHFFFAOYSA-N tin dioxide Chemical compound O=[Sn]=O XOLBLPGZBRYERU-UHFFFAOYSA-N 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
- 239000004254 Ammonium phosphate Substances 0.000 description 1
- OAICVXFJPJFONN-UHFFFAOYSA-N Phosphorus Chemical compound [P] OAICVXFJPJFONN-UHFFFAOYSA-N 0.000 description 1
- 229910000148 ammonium phosphate Inorganic materials 0.000 description 1
- 235000019289 ammonium phosphates Nutrition 0.000 description 1
- 229910052788 barium Inorganic materials 0.000 description 1
- DSAJWYNOEDNPEQ-UHFFFAOYSA-N barium atom Chemical compound [Ba] DSAJWYNOEDNPEQ-UHFFFAOYSA-N 0.000 description 1
- 229910052792 caesium Inorganic materials 0.000 description 1
- TVFDJXOCXUVLDH-UHFFFAOYSA-N caesium atom Chemical compound [Cs] TVFDJXOCXUVLDH-UHFFFAOYSA-N 0.000 description 1
- 239000004568 cement Substances 0.000 description 1
- UOUJSJZBMCDAEU-UHFFFAOYSA-N chromium(3+);oxygen(2-) Chemical class [O-2].[O-2].[O-2].[Cr+3].[Cr+3] UOUJSJZBMCDAEU-UHFFFAOYSA-N 0.000 description 1
- 239000010786 composite waste 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
- 238000005516 engineering process Methods 0.000 description 1
- 230000004907 flux Effects 0.000 description 1
- 238000005816 glass manufacturing process Methods 0.000 description 1
- 239000002927 high level radioactive waste Substances 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
- 239000007791 liquid phase Substances 0.000 description 1
- 239000010808 liquid waste Substances 0.000 description 1
- JKQOBWVOAYFWKG-UHFFFAOYSA-N molybdenum trioxide Inorganic materials O=[Mo](=O)=O JKQOBWVOAYFWKG-UHFFFAOYSA-N 0.000 description 1
- 230000007935 neutral effect Effects 0.000 description 1
- 238000009376 nuclear reprocessing Methods 0.000 description 1
- OOAWCECZEHPMBX-UHFFFAOYSA-N oxygen(2-);uranium(4+) Chemical compound [O-2].[O-2].[U+4] OOAWCECZEHPMBX-UHFFFAOYSA-N 0.000 description 1
- RVTZCBVAJQQJTK-UHFFFAOYSA-N oxygen(2-);zirconium(4+) Chemical class [O-2].[O-2].[Zr+4] RVTZCBVAJQQJTK-UHFFFAOYSA-N 0.000 description 1
- 229910052698 phosphorus Inorganic materials 0.000 description 1
- 239000011574 phosphorus Substances 0.000 description 1
- 239000002574 poison Substances 0.000 description 1
- 231100000614 poison Toxicity 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 238000000926 separation method Methods 0.000 description 1
- 239000004317 sodium nitrate Substances 0.000 description 1
- 235000010344 sodium nitrate Nutrition 0.000 description 1
- 229910001948 sodium oxide Inorganic materials 0.000 description 1
- 239000001488 sodium phosphate Substances 0.000 description 1
- 229910000162 sodium phosphate Inorganic materials 0.000 description 1
- 229910052596 spinel Inorganic materials 0.000 description 1
- 239000011029 spinel Substances 0.000 description 1
- 229910052712 strontium Inorganic materials 0.000 description 1
- CIOAGBVUUVVLOB-UHFFFAOYSA-N strontium atom Chemical compound [Sr] CIOAGBVUUVVLOB-UHFFFAOYSA-N 0.000 description 1
- RYFMWSXOAZQYPI-UHFFFAOYSA-K trisodium phosphate Chemical compound [Na+].[Na+].[Na+].[O-]P([O-])([O-])=O RYFMWSXOAZQYPI-UHFFFAOYSA-K 0.000 description 1
- 238000003826 uniaxial pressing Methods 0.000 description 1
- FCTBKIHDJGHPPO-UHFFFAOYSA-N uranium dioxide Inorganic materials O=[U]=O FCTBKIHDJGHPPO-UHFFFAOYSA-N 0.000 description 1
- 229910001928 zirconium oxide Inorganic materials 0.000 description 1
Classifications
-
- 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
- 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
-
- 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)
- Life Sciences & Earth Sciences (AREA)
- Chemical Kinetics & Catalysis (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)
- Inorganic Chemistry (AREA)
- Ceramic Engineering (AREA)
- Dispersion Chemistry (AREA)
- Processing Of Solid Wastes (AREA)
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 (20)
- 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.
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Citations (2)
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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 |
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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 (4)
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WO2009039059A1 (en) | 2007-09-20 | 2009-03-26 | Energysolutions, Llc | Mitigation of secondary phase formation during waste vitrification |
EP2195277A1 (en) * | 2007-09-20 | 2010-06-16 | Energysolutions, Llc | Mitigation of secondary phase formation during waste vitrification |
EP2195277A4 (en) * | 2007-09-20 | 2013-11-27 | Energysolutions Llc | Mitigation of secondary phase formation during waste vitrification |
US8951182B2 (en) | 2007-09-20 | 2015-02-10 | Energysolutions, Llc | Mitigation of secondary phase formation during waste vitrification |
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