WO1997025721A1 - Recovery of actinides - Google Patents

Recovery of actinides Download PDF

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Publication number
WO1997025721A1
WO1997025721A1 PCT/GB1997/000062 GB9700062W WO9725721A1 WO 1997025721 A1 WO1997025721 A1 WO 1997025721A1 GB 9700062 W GB9700062 W GB 9700062W WO 9725721 A1 WO9725721 A1 WO 9725721A1
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WIPO (PCT)
Prior art keywords
acid
plutonium
waste liquid
process according
uranium
Prior art date
Application number
PCT/GB1997/000062
Other languages
French (fr)
Inventor
Robin John Taylor
Alister Andrew Dunlop
Andrew Philip Jeapes
Ian Stewart Denniss
Peter Parkes
Jeffrey William Hobbs
Paul Richard Silverwood
Francis Robin Livens
David Collison
Original Assignee
British Nuclear Fuels Plc
Priority date (The priority date 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 date listed.)
Filing date
Publication date
Application filed by British Nuclear Fuels Plc filed Critical British Nuclear Fuels Plc
Priority to AU13899/97A priority Critical patent/AU1389997A/en
Publication of WO1997025721A1 publication Critical patent/WO1997025721A1/en

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Classifications

    • C—CHEMISTRY; METALLURGY
    • C01—INORGANIC CHEMISTRY
    • C01G—COMPOUNDS CONTAINING METALS NOT COVERED BY SUBCLASSES C01D OR C01F
    • C01G56/00—Compounds of transuranic elements
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
    • G21C19/42—Reprocessing of irradiated fuel
    • G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
    • G21C19/46—Aqueous processes, e.g. by using organic extraction means, including the regeneration of these means
    • 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
    • G21F9/04—Treating liquids
    • G21F9/06—Processing
    • G21F9/10—Processing by flocculation
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
    • Y02W30/00—Technologies for solid waste management
    • Y02W30/50—Reuse, recycling or recovery technologies

Definitions

  • This invention relates to the recovery of actinides and has particular application in the field of nuclear fuel reprocessing.
  • the invention is concerned especially, but not exclusively, with the separation of uranium, plutonium or thorium, or a mixture thereof, from other waste materials produced in the nuclear fuel reprocessing plant.
  • a mixed stream of uranium and plutonium in 30% TBP/OK is obtained following a solvent extraction which removed fission products to waste.
  • the uranium and plutonium are then separated by solvent extraction.
  • Uranium as U(VI) is extracted into TBP/OK.
  • Pu(IV) is reduced to Pu(III) by U(IV) which itself is oxidised to U(VI).
  • Pu(III) is inextractable and remains in the aqueous nitric acid phase.
  • the uranium stream is then backwashed in to aqueous nitric acid and sent to product finishing. Evaporation followed by thermal den ration is used to produce the solid uranium oxide product.
  • the plutonium stream is also purified by a further solvent extraction cycle and then sent to product finishing where it is concentrated by evaporation and precipitated as plutonium oxalate and decomposed to the oxide by calcination.
  • a process for the separation of uranium and/or plutonium from a radioactive waste liquid comprising contacting the waste liquid with a carboxylic acid thereby precipitating from the waste liquid a complex formed between the uranium and/or plutonium and the carboxylic acid, and separating the complex from the waste liquid.
  • the waste liquid is an aqueous acid solution.
  • the process of the present invention is much simpler than the above described prior art process. There is no U/Pu separation and the same reagent is used for both U and Pu precipitation. The process is flexible and U or Pu or U/Pu can be precipitated.
  • the carboxylic acid may be, for instance, an acid with a nitrogen or oxygen atom at the alpha position, preferably attached to one or more atoms other than hydrogen.
  • the carboxylic acid is a polycarboxylic acid, more preferably a polyaminocarboxylic acid.
  • Particularly preferred ligands for use with the present invention are monopicolinic acid (Hmpa), nitrilotriacetic acid (H 3 nta), dipicolinic acid (H 2 dpa), oxydiacetic acid (H 2 oda), N-(2-hydroxyethyl)iminodiacetic acid(H 3 hida), nitrilotripropionic acid (H 3 ntp) and N-(2- acetamido)iminodiacetic acid (H 2 aida).
  • Hmpa monopicolinic acid
  • nitrilotriacetic acid H 3 nta
  • dipicolinic acid H 2 dpa
  • oxydiacetic acid H 2 oda
  • nitrilotripropionic acid H 3 ntp
  • N-(2- acetamido)iminodiacetic acid H 2
  • the nature of the precipitant depends on the conditions used, including the ratio of ligand to metal.
  • the precipitation of the uranyl (UO 2 2+ ) ion is fast with H 3 nta producing an amorphous solid.
  • precipitation is relatively slow with H 2 oda and grows crystals.
  • the separated complex will then normally be treated to form the oxides by, for instance, thermal degradation.
  • the process of the present invention can be utilised to provide a novel reprocessing route.
  • the above mentioned ligands of use in the present invention can be used immediately to precipitate either Pu or a Pu/U mixture from solution.
  • This product is then thermally decomposed to the oxide.
  • the procedure provides a very fast route from fuel dissolution to final product.
  • a single reactor is used to dissolve the fuel, form the precipitate and decompose the products, thereby leading to large cost savings.
  • a preferred product of a process of the present invention is suitable after finishing for MOX fuel fabrication, either directly or after blending to the required plutonium enrichment with urania powder.
  • the precipitant is added to a concentrated nitrate stream incorporating the irradiated fuel after bulk fission products have been removed.
  • the stream then contains only uranium and is sent forward for finishing. If the stream still contains unacceptably high levels of residual plutonium, then the uranium may be finished by the ADU process at relatively low pH to leave the plutonium in solution, since plutonium precipitates at higher pH. Alternatively, the uranium could go forward for further separation/purification to meet requirements for hex conversion and enrichment.
  • the precipitant is added directly to the actinide stream to precipitate plutonium and thus eliminate the need for a separate process to achieve separation of plutonium from uranium.
  • a process involving the present invention may be represented as follows:
  • Blend down to Mox concentration with urania (if required)
  • Condition powder for pressing (not required if urania is free flowing, eg AUC)
  • New Mox fuel In this process the irradiated fuel is first subjected to dissolution and bulk fission products are extracted. The plutonium concentration is then measured and, if required, the solution is treated to form the appropriate oxidation state. A stoichiometric (based on total uranium and plutonium) amount of precipitant, is added. A precipitate is produced which contains all the plutonium and a proportion of the uranium. This precipitate is filtered out and the uranium stream is passed on for further purification if required. The precipitate is thermally decomposed to oxide and the resultant oxide is blended down to Mox concentration with urania. The powder is then conditioned for pressing if not free flowing.
  • the final stages involve the pressing of the powder into pellets and the sintering of the pellets to produce a Mox fuel.
  • This example has the particular advantage that plutonium is never separated from uranium.
  • the use of co-precipitated material in fact benefits MOX fuel production by reducing powder processing stages during fabrication.
  • the organic acid was dissolved in water (Hmpa, H 2 oda at 20°C and H 3 nta, H 3 hida, H 2 dpa, H 2 aida at 80°C) and added to a solution containing actinide (VI) or actinide (IV) ions.
  • the pH of the solution was then adjusted, if necessary, with a base (NaOH, K 2 CO 3 or NaHCO 3 ) until precipitation was initiated.
  • Complex formation was then followed by vacuum/oven drying and analysis of the products. In some cases the composition of the product depends on the stoichiometry of reaction. Thermogravimetric analysis was employed to obtain the temperature of thermal degradation to the metal oxide.
  • Th Due to difficulties in handling Pu, Th has been used as an analogue for Pu. This is acceptable since both are actinide elements which form ions of oxidation state +4 in solution. Since complex formation in tetravalent actinide ions is known to increase in the order Th + ⁇ l + ⁇ Np 4+ ⁇ Pu 4+ , better results with Pu than Th are likely to be achieved.
  • Reaction times vary from immediate precipitation to gradual growth of the products over a number of hours.

Abstract

A process for the separation of uranium and/or plutonium from a radioactive waste liquid comprising contacting the waste liquid with a carboxylic acid with precipitates from the waste liquid a complex formed between the uranium and/or plutonium and the carboxylic acid. The complex is then separated from the waste liquid.

Description

RECOVERY OF ACTINIDES
FIELD OF THE INVENTION
This invention relates to the recovery of actinides and has particular application in the field of nuclear fuel reprocessing. The invention is concerned especially, but not exclusively, with the separation of uranium, plutonium or thorium, or a mixture thereof, from other waste materials produced in the nuclear fuel reprocessing plant.
BACKGROUND OF THE INVENTION
In a known process, a mixed stream of uranium and plutonium in 30% TBP/OK is obtained following a solvent extraction which removed fission products to waste. The uranium and plutonium are then separated by solvent extraction. Uranium as U(VI) is extracted into TBP/OK. Pu(IV) is reduced to Pu(III) by U(IV) which itself is oxidised to U(VI). Pu(III) is inextractable and remains in the aqueous nitric acid phase. The uranium stream is then backwashed in to aqueous nitric acid and sent to product finishing. Evaporation followed by thermal den ration is used to produce the solid uranium oxide product.
The plutonium stream is also purified by a further solvent extraction cycle and then sent to product finishing where it is concentrated by evaporation and precipitated as plutonium oxalate and decomposed to the oxide by calcination.
These known processes involve complex procedures with large plant sizes and costs.
STATEMENTS OF INVENTION
According to the present invention there is provided a process for the separation of uranium and/or plutonium from a radioactive waste liquid comprising contacting the waste liquid with a carboxylic acid thereby precipitating from the waste liquid a complex formed between the uranium and/or plutonium and the carboxylic acid, and separating the complex from the waste liquid.
Preferably the waste liquid is an aqueous acid solution.
The process of the present invention is much simpler than the above described prior art process. There is no U/Pu separation and the same reagent is used for both U and Pu precipitation. The process is flexible and U or Pu or U/Pu can be precipitated. The carboxylic acid may be, for instance, an acid with a nitrogen or oxygen atom at the alpha position, preferably attached to one or more atoms other than hydrogen. Preferably, the carboxylic acid is a polycarboxylic acid, more preferably a polyaminocarboxylic acid.
Particularly preferred ligands for use with the present invention are monopicolinic acid (Hmpa), nitrilotriacetic acid (H3nta), dipicolinic acid (H2dpa), oxydiacetic acid (H2oda), N-(2-hydroxyethyl)iminodiacetic acid(H3hida), nitrilotripropionic acid (H3ntp) and N-(2- acetamido)iminodiacetic acid (H2aida).
The formulae for these preferred liquids are given in Figure 1 of the accompanying drawings.
DETAILED DESCRIPTION OF INVENTION
The nature of the precipitant depends on the conditions used, including the ratio of ligand to metal. The precipitation of the uranyl (UO2 2+) ion is fast with H3nta producing an amorphous solid. By contrast precipitation is relatively slow with H2oda and grows crystals.
Having precipitated the complex from the waste liquid, the separated complex will then normally be treated to form the oxides by, for instance, thermal degradation.
The precipitation of Pu(IV) takes place preferentially over U(VI) (present as a hexavalent dioxo-actinyl ion) due to the higher charge density of the Pu ion. As a result, by controlling conditions such as ligand concentration, it is possible to obtain reasonably accurate dilution of the Pu product with U in the formation of a MOX fuel or a MOX fuel precursor.
The process of the present invention can be utilised to provide a novel reprocessing route. After fuel dissolution in nitric acid, and adjustment of the acidity, the above mentioned ligands of use in the present invention can be used immediately to precipitate either Pu or a Pu/U mixture from solution. This product is then thermally decomposed to the oxide. The procedure provides a very fast route from fuel dissolution to final product. In a particular embodiment of the present invention, a single reactor is used to dissolve the fuel, form the precipitate and decompose the products, thereby leading to large cost savings. A preferred product of a process of the present invention is suitable after finishing for MOX fuel fabrication, either directly or after blending to the required plutonium enrichment with urania powder.
Typically, the precipitant is added to a concentrated nitrate stream incorporating the irradiated fuel after bulk fission products have been removed. The stream then contains only uranium and is sent forward for finishing. If the stream still contains unacceptably high levels of residual plutonium, then the uranium may be finished by the ADU process at relatively low pH to leave the plutonium in solution, since plutonium precipitates at higher pH. Alternatively, the uranium could go forward for further separation/purification to meet requirements for hex conversion and enrichment.
Alternatively, the precipitant is added directly to the actinide stream to precipitate plutonium and thus eliminate the need for a separate process to achieve separation of plutonium from uranium.
A process involving the present invention may be represented as follows:
Fuel
Head end dissolution
U+Pu+fps
Primary extraction of bulk fps to HA stream
U+Pu
Measure Pu concentration
Condition oxidation states if required
Add a stoichiometric amount of precipitant X
Figure imgf000005_0001
Filter out precipitant, pass U stream on for further purification if required
(U,Pu)X
Thermally decompose to oxide as at present
OLEuilB
Blend down to Mox concentration with urania (if required) Condition powder for pressing (not required if urania is free flowing, eg AUC)
Press fee
Press pellets Sinter pellets
New Mox fuel In this process the irradiated fuel is first subjected to dissolution and bulk fission products are extracted. The plutonium concentration is then measured and, if required, the solution is treated to form the appropriate oxidation state. A stoichiometric (based on total uranium and plutonium) amount of precipitant, is added. A precipitate is produced which contains all the plutonium and a proportion of the uranium. This precipitate is filtered out and the uranium stream is passed on for further purification if required. The precipitate is thermally decomposed to oxide and the resultant oxide is blended down to Mox concentration with urania. The powder is then conditioned for pressing if not free flowing. The final stages involve the pressing of the powder into pellets and the sintering of the pellets to produce a Mox fuel. This example has the particular advantage that plutonium is never separated from uranium. The use of co-precipitated material in fact benefits MOX fuel production by reducing powder processing stages during fabrication.
A process of the present invention has the following advantageous features:
1. Reactions take place in aqueous solution.
2. Timescale for reactions is short.
3. Cheap, uncomplicated, precipitants are used.
4. Small amounts of precipitant, with respect to actinide, are used.
5. Yields of complex are high.
6. Complexes readily undergo thermal degradation to the oxides of the actinides.
7. Insolubility of the products provides a very good reaction driving force.
EXAMPLES OF THE INVENTION Experimental procedure
The organic acid was dissolved in water (Hmpa, H2oda at 20°C and H3nta, H3hida, H2dpa, H2aida at 80°C) and added to a solution containing actinide (VI) or actinide (IV) ions. The pH of the solution was then adjusted, if necessary, with a base (NaOH, K2CO3 or NaHCO3) until precipitation was initiated. Complex formation was then followed by vacuum/oven drying and analysis of the products. In some cases the composition of the product depends on the stoichiometry of reaction. Thermogravimetric analysis was employed to obtain the temperature of thermal degradation to the metal oxide. Techniques used to characterise the complexes included infrared and nuclear magnetic resonance spectroscopy and (in the cases of crystalline products) x-ray crystallography. The formula of hexamethylenetetramine (hmta), a counter-ion and in situ base, is given in Figure 1.
The results are shown on the following pages. Due to difficulties in handling Pu, Th has been used as an analogue for Pu. This is acceptable since both are actinide elements which form ions of oxidation state +4 in solution. Since complex formation in tetravalent actinide ions is known to increase in the order Th + < l + < Np4+ < Pu4+, better results with Pu than Th are likely to be achieved.
UO22+ :L Hmpa H2dpa H7oda H.hida H3nta
1 : 1 no base no base no base hmta base/ion no base ppt 30 mins ppt 30 mins ppt 120 mins ppt 2 days ppt 10 mins yellow crystalline yellow crystalline yellow crystalline yellow amorphous yellow amorpho 95% yield 95% yield 98% yield (same as 100B 99% yield [UO2(dpa)(H2O)]n [UO2(oda)]n [UO2(hida)(H2O)] [UO2(Hnta)
483°C decompose 525°C decompose 390°C decompose [Hhmta] (H2O)2].3H2O 510°C decompos
1 :2 no base no base no base hmta base/ion no base ppt<l min ppt 120 mins ppt 120 mins ppt 2 days ppt 10 mins yellow fluorescent green crystalline yellow crystalline yellow amorphous yellow amorpho amorphous 95% yield 95% yield (same as 98B) 95% yield
95% yield [UO2(Hdpa)2].4H2O [UO2(Hoda)2] [UO2(hida)(H2O)J [UO2(H2nta)2]
[UO2(mpa)2] 527°C decompose [Hhmta] 3H2O
482°C decompose 512°C decompos soluble: dmf, dmso insol:
H2O, MeOH
Th 4"+ :L Hmpa H2dpa H2oda H^hida H3nta
2:1 no product no base no product K2CO3base no product very soluble? white crystalline white amorphous
Th2C7H19N9O28 Th2C6H20N10O52
494 °C decompose
1 : 1 base and anion no base no base NaOH base no product yellow 'paste' white crystalline white crystalline no product forced precipitation [Th(dpa)2(H2O)4] [Th(oda)2[H2O)4] .nH2O unworkable product ThC14H14N2O12 ThCgH27O21
532 °C decompose
1 :2 no product no base no base no product complex produc white crystalline white crystalline hmta base / ion
[Th(dpa)2(H2O)4] [Th(oda)2(H2O)4] .nH2O ThC16H15N4O10
ThC,4H,4N2O12 ThC8H27O21 Soluble in wate
536 °C decompose 484 C decompose
1 :3 HCO3- base hmta base / ion no base no product no product white crystals white crystalline white crystalline soluble in water [Th(dpa)3]2 [Hhmta] .3H2O [Th(oda)2(H2O)4] .nH2O ThC25H27N10 ThC33H41Nl lO15 ThC8H27O21 [Th(mpa)4] 535 °C decompose same as 62B .xH2O.yNa zNO3
As will be appreciated from the above results, a range of actinide polycarboxylate complexes (some monomeric and others polymeric) have been prepared and characterised. Thermal degradation of these compounds occurs within the temperature range 390-536°C and calculations suggest that the decomposition products are the metal oxides.
Reaction times vary from immediate precipitation to gradual growth of the products over a number of hours.
The reaction yields have been estimated as very high. H3nta was observed to be the most efficient precipitant of UO2 \ whilst H2dpa generally leads to clean, crystalline products on which we concentrated further study and spectroscopic characterisation. The use of hmta as an in situ base and counter-ion has allowed anionic species to be brought out of solution.
The very low solubility of most of these compounds in most common solvents has proven to be a good reaction driving force and has made separation of the products from other (more soluble) contaminants quite feasible.

Claims

1. A process for the separation of uranium and/or plutonium from a radioactive waste liquid comprising contacting the waste liquid with a carboxylic acid thereby precipitating from the waste liquid a complex formed between the uranium and/or plutonium and the carboxylic acid, and separating the complex from the waste liquid.
2. A process according to claim 1 wherein the waste liquid is an aqueous acid solution.
3. A process according to claim 1 or 2 wherein the carboxylic acid has a nitrogen or oxygen
atom at the alpha position.
4. A process according to claim 3 wherein the nitrogen or oxygen atom is attached to one or
more atoms other than hydrogen.
5. A process according to any preceding claims wherein the carboxylic acid is a
polycarboxylic acid.
6. A process wherein according to claim 1 carboxylic acid is monopicolinic acid (HmpA),
nitrilotriacetic acid (H5nta), dipicolinic acid (H2dpa), oxydiacetic acid (H2oda), N-(2- hydroxyethyl) iminodiacetic acid (H3hida), nitrilotripropionic acid (H3ntp) or N-(2-
acetamido)iminodiacetic acid (H2aida).
7. A process according to any of the preceding claims wherein the radioactive waste liquid includes both uranium and plutonium and the carboxylic acid is capable of precipitating from the
solution at least some of the plutonium with at least some of the uranium.
8. A process according to claim 7 wherein substantially the whole of the plutonium is
precipitated from the solution.
9. A process according to any of the preceding claims wherein the waste liquid is an
aqueous nitrate solution derived from irradiated fuel from a nuclear process.
10. A process according to any of the preceding claims wherein the product is suitable after finishing for MOX fuel fabrication, either directly or after blending to the required plutonium
enrichment with urania powder.
0 0
Figure imgf000013_0001
H, hi da
Figure imgf000013_0002
0
Figure imgf000013_0003
H„dpa hmta
1
Figure imgf000013_0004
PCT/GB1997/000062 1996-01-08 1997-01-10 Recovery of actinides WO1997025721A1 (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
AU13899/97A AU1389997A (en) 1996-01-08 1997-01-10 Recovery of actinides

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
GB9600286A GB9600286D0 (en) 1996-01-08 1996-01-08 Recovery of actinides
GB9600286.0 1996-01-10

Publications (1)

Publication Number Publication Date
WO1997025721A1 true WO1997025721A1 (en) 1997-07-17

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GB (1) GB9600286D0 (en)
WO (1) WO1997025721A1 (en)
ZA (1) ZA97161B (en)

Cited By (4)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
WO1999066086A1 (en) * 1998-06-19 1999-12-23 Commissariat A L'energie Atomique Method for dissolving plutonium or a plutonium alloy and converting it into nuclear fuel
WO2002028778A1 (en) * 2000-10-05 2002-04-11 Commissariat A L'energie Atomique Method for co-precipitating actinides and method for preparing mixed actinide oxides
US7829043B2 (en) 2004-05-28 2010-11-09 Commissariat A L'energie Atomique Method for coprecipitation of actinides in different oxidation states and method for preparation of mixed compounds of actinides
CN103760273A (en) * 2014-01-28 2014-04-30 中国原子能科学研究院 Analysis method for trace oxalic acid root in mother liquor of plutonium oxalate precipitation

Citations (4)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
GB991418A (en) * 1961-10-11 1965-05-05 Europ Pour Le Traitement Chimi Process for separating pure plutonium, thorium and uranium from impurities
GB1087322A (en) * 1964-03-20 1967-10-18 Atomic Energy Authority Uk Recovery of uranium and/or plutonium from solution
US4358426A (en) * 1980-12-17 1982-11-09 The United States Of America As Represented By The United States Department Of Energy Method for cleaning solution used in nuclear fuel reprocessing
JPH0815483A (en) * 1994-06-24 1996-01-19 Ishikawajima Harima Heavy Ind Co Ltd Method for solvent extraction of transuranium elements

Patent Citations (4)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
GB991418A (en) * 1961-10-11 1965-05-05 Europ Pour Le Traitement Chimi Process for separating pure plutonium, thorium and uranium from impurities
GB1087322A (en) * 1964-03-20 1967-10-18 Atomic Energy Authority Uk Recovery of uranium and/or plutonium from solution
US4358426A (en) * 1980-12-17 1982-11-09 The United States Of America As Represented By The United States Department Of Energy Method for cleaning solution used in nuclear fuel reprocessing
JPH0815483A (en) * 1994-06-24 1996-01-19 Ishikawajima Harima Heavy Ind Co Ltd Method for solvent extraction of transuranium elements

Non-Patent Citations (1)

* Cited by examiner, † Cited by third party
Title
PATENT ABSTRACTS OF JAPAN vol. 96, no. 005 *

Cited By (9)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
WO1999066086A1 (en) * 1998-06-19 1999-12-23 Commissariat A L'energie Atomique Method for dissolving plutonium or a plutonium alloy and converting it into nuclear fuel
FR2779974A1 (en) * 1998-06-19 1999-12-24 Commissariat Energie Atomique PROCESS FOR THE DISSOLUTION OF PLUTONIUM OR A PLUTONIUM ALLOY
US6623712B1 (en) 1998-06-19 2003-09-23 Commissariat A L'energie Atomique Method for dissolving plutonium or a plutonium alloy and converting it into nuclear fuel
WO2002028778A1 (en) * 2000-10-05 2002-04-11 Commissariat A L'energie Atomique Method for co-precipitating actinides and method for preparing mixed actinide oxides
FR2815035A1 (en) * 2000-10-05 2002-04-12 Commissariat Energie Atomique ACTINIDE CO-PRECIPITATION PROCESS AND ACTINIDE MIXED OXIDE PREPARATION PROCESS
US7169370B2 (en) 2000-10-05 2007-01-30 Commissariat A L'energie Atomique Method for co-precipitating actinides and method for preparing mixed actinide oxides
US7829043B2 (en) 2004-05-28 2010-11-09 Commissariat A L'energie Atomique Method for coprecipitation of actinides in different oxidation states and method for preparation of mixed compounds of actinides
CN103760273A (en) * 2014-01-28 2014-04-30 中国原子能科学研究院 Analysis method for trace oxalic acid root in mother liquor of plutonium oxalate precipitation
CN103760273B (en) * 2014-01-28 2014-11-19 中国原子能科学研究院 Analysis method for trace oxalic acid root in mother liquor of plutonium oxalate precipitation

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AU1389997A (en) 1997-08-01
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