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WO1999023668A1 - Retraitement des combustibles nucleaires - Google Patents

Retraitement des combustibles nucleaires Download PDF

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Publication number
WO1999023668A1
WO1999023668A1 PCT/GB1998/003236 GB9803236W WO9923668A1 WO 1999023668 A1 WO1999023668 A1 WO 1999023668A1 GB 9803236 W GB9803236 W GB 9803236W WO 9923668 A1 WO9923668 A1 WO 9923668A1
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WO
WIPO (PCT)
Prior art keywords
solvent
unit
iii
aqueous
fed
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.)
Ceased
Application number
PCT/GB1998/003236
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English (en)
Inventor
Emmanuel Thomas Gaubert
Gavin Paul Towler
Andrew Lindsay Wallwork
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Sellafield Ltd
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
Publication of WO1999023668A1 publication Critical patent/WO1999023668A1/fr
Anticipated expiration legal-status Critical
Ceased legal-status Critical Current

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Classifications

    • 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
    • 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 nuclear fuel reprocessing and is particularly concerned with the separation of uranium from plutonium and neptunium.
  • the organic phase is subjected to separation of fission products by solvent extraction and typically then to separation of technetium, before the so-called U/Pu split.
  • U/Pu split Pu(IV) is reduced to Pu(III) which is inextractable into the organic phase and therefore follows the aqueous stream while the U remains in the organic stream.
  • the reducing agent used in the U/Pu split is U(IV).
  • Np(VI) in the solvent stream is also reduced by the U(IV) to Np(IV).
  • Np(IV) is extractable into the solvent and so exits the contactor in the solvent stream with the U product.
  • Hydrazine nitrate is normally used to stabilise the U(IV) and Pu(III) against oxidation by, in particular, HNO 2 .
  • the unit for carrying out the partitioning of the U and Pu in practice comprises a contactor having a multiplicity of stages, for example six stages might be used in a modern centrifugal contactor.
  • Np is not separated from U so additional downstream processes are needed to remove Np from U .
  • Neptunium valency control can be a significant problem in Purex reprocessing.
  • Neptunium is present in the Purex process as a mixture of three different valence states: Np(IV), (V) and (VI).
  • Np(IV) and (VI) are both extractable into the solvent phase whereas Np(V) is inextractable into this phase.
  • Np is normally stabilised in the (V) oxidation state. This is a complex matter, since not only is it the middle oxidation state of three but Np(V) also undergoes competing reactions, such as disproportionation to Np(IV) and (VI) and is oxidised to Np(VI) by nitric acid.
  • Neptunium control is therefore difficult and efficient neptunium control is a major aim of an advanced reprocessing programme.
  • the present invention provides a spent fuel reprocessing method in which an organic phase containing U, Pu and Np is contacted with a reductant to reduce Pu to Pu(III) and Np(VI) to Np(V), and the Pu(III) and Np(V) are backwashed into a first aqueous phase and the treated solvent phase is contacted with a hydrophilic complexant for forming a complex with Np(IV), which is backwashed into a second aqueous phase to scrub the Np(IV) from the solvent phase in which the U remains.
  • the organic phase is contacted with the reductant and the Pu(III) plus Np(V) are backwashed in a first contactor unit from which the organic phase is fed to a second contactor unit in which the Np(IV) is complexed and backwashed.
  • the two aqueous products are combined and fed into a contactor for recovery of uranium into an organic phase which is mixed with first organic active feed and fed to the first multistage contactor.
  • the contactors are normally multi-stage contactors.
  • a spent fuel reprocessing method which effectively routes Np to an aqueous solution, independent of its initial oxidation state or states. It is characterised in that hydroxylamine is used to reduce any Np(VI) to Np(V) and in that formohydroxamic acid is subsequently used to form a complex with Np(IV) and to reduce any residual Np(VI), whereby routing all the neptunium present into the aqueous phase during solvent extraction.
  • the invention includes a Purex reprocessing plant in which there are arranged in series along a solvent stream flowpath (i) a unit for extraction of uranium into the solvent from an aqueous phase, (ii) a unit for treating the solvent stream when combined with a solvent stream containing U, Pu and Np with a reductant to reduce Pu(IV) to Pu(III) and Np(VI) to Np(V) and for backwashing the Pu(III) and Np(V) into an aqueous phase which is then fed to the unit (i), and (iii) a unit for contacting the solvent stream with a complexant for forming a water-soluble complex with Np(IV) and for backwashing the complex into an aqueous phase which is then fed to the unit (i).
  • the invention also provides a spent fuel reprocessing method in which a solvent stream passes in series through the aforesaid units.
  • Figure 1 is a partial flowsheet of a Purex reprocessing process incorporating the methods of the invention.
  • Figure 1 is therefore a flowsheet of part of a Purex reprocessing plant. The following symbols are used in the Figure:
  • the flowsheet contains the units shown in Table 1. Table 1. Units used in the Purex reprocessing plant in Figure 1.
  • nitric acid solution resulting from dissolution of the spent fuel is subject to removal of fission products and normally Tc, for example a conventional manner.
  • the resulting organic stream containing U, Pu, Np and, in some cases, Tc, is sent to the U/Pu split operation where it is reduced; in preferred embodiments the reductant is hydroxylamine (HAN).
  • intermediate solvent (organic) stream 12 of Figure 1 is sent to unit BX of the apparatus illustrated in Figure 1.
  • the aqueous feeds, intermediate solvent streams and product streams shown in Figure 1 in relation to the U/Pu split and Np rejection operations are as follows:
  • the organic stream 12 is contacted with HAN in the unit BX, which is a multi-stage contactor in the illustrated embodiment.
  • the HAN reduces Np(VI) to Np(V), which is inextractable into the organic phase, and it reacts with Pu(IV) to give inextractable Pu(III).
  • the organic phase loaded with U, Np(IV) and residual Pu(IV) goes from unit BX to unit NpS, in this case a multi-stage contactor unit, where a polish of neptunium decontamination is performed.
  • Formohydroxamic acid (FHA) may be used to reduce/ complex the Np. as described in WO 97/30456.
  • FHA Formohydroxamic acid
  • Np is removed from the uranium product solvent stream using FHA as a complexant for Np(IV) and a reductant for any residual Np(VI).
  • Any residual Pu(IV) in this contactor will also be removed from the solvent stream by complexation with FHA.
  • the contactor is operated at room temperature to minimise FHA hydrolysis, but this is not an essential requirement.
  • the aqueous product of NpS is sent directly to unit BS to recover uranium therefrom.
  • Unit BX where Pu and Np are reduced is bypassed.
  • Unit BS is suitably a multi-stage contactor in which the uranium is re-extracted from the aqueous stream into solvent.
  • the method of the invention dispenses with the separation of Pu and Np, which is used in commercial reprocessing plants. Accordingly, the plant may be smaller and the solvent and aqueous flows are reduced, resulting in both environmental and economic benefits.
  • the method features excellent Np control (U, Np separation) in using, in preferred embodiments, both HAN and FHA to reduce/complex Np. Both Pu and Np may be efficiently separated from the U-loaded solvent stream.
  • a yet further benefit of preferred methods of the invention is that no U(IV) is used as a reductant Therefore, no U(IV) is backwashed with the Pu, Np product, which is thus purer. The process gives an opportunity for the number of stages in the U/Pu split operation to be decreased. Moreover, no depleted U(IV) is added to the 235U to be recovered and the final U stream is therefore more suitable for a uranium enrichment process.
  • Tc separation may be dispensed with if a low Tc specification is acceptable for the Pu, Np product and U product.
  • the Decontamination Factor (DF) of an operation is calculated as feed molar flowrate divided by the outlet molar flowrate. The DF values and U, Pu and Np concentrations were determined for a simulated operation of a reprocessing plant incorporating the invention. Satisfactory results were obtained.
  • the above-described process exemplifies a Purex reprocessing method, in which the active solvent feed entering the U/Pu split operation is treated to reduce Pu(IV) to Pu(III) and Np(VI) to Np(V). Those reduced species are backwashed into an aqueous stream and the treated solvent stream is fed to a neptunium polishing unit to backwash remaining Np(IV) into another aqueous stream.
  • the two aqueous streams are fed without intermediate treatment to a uranium recovery unit to extract uranium into a solvent stream, the Pu and Np remaining together in the aqueous stream.
  • the invention thus enables the production of a Pu, Np product from nuclear reprocessing. This is beneficial because Np is a "burnable" neutron poison and if the Pu is reused as a fuel it does not matter if Np is present. Furthermore it is an advantage to produce impure Pu products in that it prevents proliferation of nuclear weapons. Finally, it is better to remove Np with Pu than with U because U is not very radioactive and Np would be a radioactive contaminate.
  • Uranium and/or plutonium recovered using a method of the invention may be formed into fissile material, for example a fuel pellet.
  • exemplary fissile material is MOX fuel.
  • the invention therefore includes a process for reprocessing nuclear fuel to form a fissile material optionally in the form of a fuel pellet, a fuel pin or a fuel assembly, the process comprising performing a method of the invention.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Inorganic Compounds Of Heavy Metals (AREA)

Abstract

Cette application concerne le retraitement de combustibles nucléaires et plus particulièrement la séparation d'uranium du plutonium et du neptunium. Le procédé comprend le retraitement de combustible épuisé dans lequel une phase organique contenant U, Pu et Np est mise en contact avec un agent réducteur pour réduire Pu en Pu(III) et Np(VI) en Np(V). Pu(III) et Np(V) sont extraits sous forme de phase aqueuse alors que Np(IV) reste dans une phase solvant. La phase solvant est mise en contact avec un agent complexant hydrophile qui forme un complexe avec Np(IV). Le complexe est ensuite extrait sous forme d'une autre phase aqueuse. U reste dans la phase solvant à partir de laquelle on peut l'isoler. On décrit également une usine de retraitement du type Purex qui comprend un appareil pour séparer l'uranium du plutonium et du neptunium.
PCT/GB1998/003236 1997-10-31 1998-10-29 Retraitement des combustibles nucleaires Ceased WO1999023668A1 (fr)

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
GB9722927.2 1997-10-31
GB9722927A GB9722927D0 (en) 1997-10-31 1997-10-31 Nuclear fuel reprocessing

Publications (1)

Publication Number Publication Date
WO1999023668A1 true WO1999023668A1 (fr) 1999-05-14

Family

ID=10821318

Family Applications (1)

Application Number Title Priority Date Filing Date
PCT/GB1998/003236 Ceased WO1999023668A1 (fr) 1997-10-31 1998-10-29 Retraitement des combustibles nucleaires

Country Status (2)

Country Link
GB (1) GB9722927D0 (fr)
WO (1) WO1999023668A1 (fr)

Cited By (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
WO2000013188A1 (fr) * 1998-08-28 2000-03-09 British Nuclear Fuels Plc Traitement de combustible nucleaire consistant a reduire np(vi) en np(v) au moyen d'une hydroxylamine hydrophile substituee
US6413482B1 (en) 1998-08-28 2002-07-02 British Nuclear Fuels Plc Method for reprocessing nuclear fuel by employing oximes
FR2880180A1 (fr) * 2004-12-29 2006-06-30 Cogema Perfectionnement du procede purex et ses utilisations

Citations (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US4229421A (en) * 1977-09-16 1980-10-21 British Nuclear Fuels Limited Purification of plutonium
WO1996011477A1 (fr) * 1994-10-05 1996-04-18 British Nuclear Fuels Plc Traitement des liquides
WO1997030456A1 (fr) * 1996-02-14 1997-08-21 British Nuclear Fuels Plc Retraitement de combustible nucleaire

Patent Citations (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US4229421A (en) * 1977-09-16 1980-10-21 British Nuclear Fuels Limited Purification of plutonium
WO1996011477A1 (fr) * 1994-10-05 1996-04-18 British Nuclear Fuels Plc Traitement des liquides
WO1997030456A1 (fr) * 1996-02-14 1997-08-21 British Nuclear Fuels Plc Retraitement de combustible nucleaire

Cited By (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
WO2000013188A1 (fr) * 1998-08-28 2000-03-09 British Nuclear Fuels Plc Traitement de combustible nucleaire consistant a reduire np(vi) en np(v) au moyen d'une hydroxylamine hydrophile substituee
US6413482B1 (en) 1998-08-28 2002-07-02 British Nuclear Fuels Plc Method for reprocessing nuclear fuel by employing oximes
US6444182B1 (en) * 1998-08-28 2002-09-03 British Nuclear Fuels Plc Nuclear fuel reprocessing using hydrophilic substituted hydroxylamines
FR2880180A1 (fr) * 2004-12-29 2006-06-30 Cogema Perfectionnement du procede purex et ses utilisations
WO2006072729A1 (fr) * 2004-12-29 2006-07-13 Compagnie Generale Des Matieres Nucleaires Perfectionnement du procede purex et ses utilisations
US7731870B2 (en) 2004-12-29 2010-06-08 Compagnie General Des Matieres Nucleaires Purex method and its uses

Also Published As

Publication number Publication date
GB9722927D0 (en) 1998-01-07

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