WO1999023668A1

WO1999023668A1 - Nuclear fuel reprocessing

Info

Publication number: WO1999023668A1
Application number: PCT/GB1998/003236
Authority: WO
Inventors: Emmanuel Thomas Gaubert; Gavin Paul Towler; Andrew Lindsay Wallwork
Original assignee: British Nuclear Fuels Plc
Priority date: 1997-10-31
Filing date: 1998-10-29
Publication date: 1999-05-14
Also published as: GB9722927D0

Abstract

This application relates to nuclear fuel reprocessing and is particularly concerned with the separation of uranium from plutonium and neptunium. The method provides spent fuel reprocessing 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). The Pu(III) and Np(V) are extracted into an aqueous phase with Np(VI) remaining in a solvent phase. The solvent phase is contacted with a hydrophilic complexant which form a complex with Np(IV). The complex is then extracted into a further aqueous phase. The U remains in the solvent phase from which it can be isolated. There is also described a Purex reprocessing plant which comprises apparatus for separation of uranium from plutonium and neptunium.

Description

NUCLEAR FUEL REPROCESSING

This invention relates to nuclear fuel reprocessing and is particularly concerned with the separation of uranium from plutonium and neptunium.

Most commercial reprocessing plants use the Purex process, in which the spent fuel is dissolved in nitric acid and the dissolved uranium and plutonium are subsequently extracted from the nitric acid solution into an organic phase of tributyl phosphate (TBP) dissolved in an inert hydrocarbon such as odourless kerosene. The organic phase is then subjected to solvent extraction techniques to partition the uranium from the plutonium.

More particularly, 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. In the 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. Usually, 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, HNO₂. 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.

There are disadvantages with such a process: • hydrazine is catalytically decomposed by Tc(VII) ions

• hydrazine can form undesirable oxidation products (azides, ammonium salts)

• U(IV) must be produced in a separate process on plant, thus increasing costs

• two reagents are needed

• Np is not separated from U so additional downstream processes are needed to remove Np from U .

It is also a disadvantage of current commercial Purex processes that they use a three cycle flowsheet [(1) the so-called HA cycle in which fission products are separated and the U/Pu split is performed; (2) the UP cycle in which the uranium stream is purified; (3) the PP cycle in which the plutonium stream is purified]. It is therefore desired to provide an Advanced Purex process in which there is a single solvent extraction cycle.

Moreover, 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. In order to direct Np to raffmate streams, 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.

In a preferred class of methods, 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.

Also provided is 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.

The present invention is further described by way of example only with reference to the accompanying drawings, in which:

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:

Ai = Aqueous Feeds Si = Solvent feeds Ii = Intermediate solvent streams Pi = Product streams Double arrows = Solvent streams Single arrows = Aqueous streams

The flowsheet contains the units shown in Table 1. Table 1. Units used in the Purex reprocessing plant in Figure 1.

In preferred embodiments of the invention, therefore, 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). More particularly, 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:

A6: hydroxylamine

A7: formohydroxamic acid

12: solvent product of Tc removal

13 : U-containing product of Np/Pu removal

P3: Np/Pu product

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. Specifically, 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. Generally, 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 Pu, Np product may either be disposed of or used in the manufacture of MOX fuel (MOX = Mixed Oxide (U + Pu)).

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.

Conventional Purex processes include, after fission product separation, a Tc separation operation. The reason for this is that the hydrazine nitrate customarily used to stabilise U(IV) and Pu(III) takes part in autocatalytic reactions with Tc. In preferred methods of the present invention, no U(IV) is added. Moreover, hydroxylamine, which reacts with Tc only very slowly, acts as a nitrous acid scavenger and so reduces Pu(III) re-oxidation. Accordingly, 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.

It will be appreciated that 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.

Claims

1. 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), 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, with Np(IV), a complex which is backwashed into a second aqueous phase, to scrub the Np(IV) from the solvent phase in which the U remains.

2. A method of claim 1 wherein the reductant is hydroxylamine.

3. A method of claim 1 or claim 2 wherein the complexant is formohydroxamic acid.

4. A method of any of claims 1 to 3 wherein 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 being combined and fed into a contactor for re- extraction of uranium into an organic phase which is mixed with the organic active feed and fed to the first contactor.

5. A Purex reprocessing method, characterised in that the active solvent feed entering the U/Pu split operation is treated to reduce Pu to Pu(III) and Np(VI) to Np(V) which reduced species are backwashed into an aqueous stream, the treated solvent feed being fed to a neptunium polishing unit to backwash the remaining Np(IV) into another aqueous stream and the two aqueous streams being fed without intermediate treatment to a uranium extraction unit to re-extract uranium into a solvent stream.

6. A spent fuel reprocessing method which routes Np to an aqueous solution irrespective of its initial oxidation state or states, 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).

7. 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 any of claims 1 to 6.

8. 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 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 without intermediate treatment to the unit (i).