The present invention relates to the processing of irradiated oxide nuclear fuel material.

Nuclear fuels, discharged after irradiation in a nuclear

power reactor, are currently reprocessed by dissolution in nitric acid followed by one or more cycles of solvent extraction to separate the uranium and plutonium from radioactive fission products, minor actinides and other

impurities, and from each other. This prior art process yields very pure products but is expensive in capital and operating costs, and produces large volumes of aqueous waste of which the treatment and disposal account for a substantial proportion of the cost. Furthermore, the purity of the end product of plutonium nitrate or oxide arouses objections in some quarters as an alleged potential route to the proliferation of nuclear weapons due to the purity of the final product and the relative ease with which it may be handled.

Such high levels of purification are currently necessary to

permit the direct handling involved in the current routes

for recycling irradiated uranium or plutonium into new

fuel, or to avoid the contamination of equipment designed

to process virgin uranium.

Other prior art methods of reprocessing irradiated fuel

involve pyrochemicai methods where dissolution of the fuel

is achieved by molten salt techniques. The irradiated fuel

is dissolved into molten alkali metal halide salts such as

chlorides or fluorides . This method leaves some of the

radioactive fission products in the final processed

material, but this can be an advantage in that it makes the

material easier to trace if diverted and much more

difficult to handle thereafter, but relatively easily

handled in plants properly designed and equipped to do so.

Furthermore, the small residual proportion of radioactive

fission products is not in the least detrimental in the

eventual fuel rods which will be produced from the

reprocessed material as these would, in any case, be formed

very quickly after the fuel went into operation in a

reactor. However, the real disadvantage in the prior art

salt processing techniques is that the molten salt

materials are highly corrosive to plant and waste containers and are also, in the case of fluorides, highly

toxic.

A further disadvantage of the prior art salt processing

techniques is that the halide salts allow the residual entrained metal ions to be relatively mobile in waste

burial sites in case of leaching.

It is an object of the present invention to provide a method for the reprocessing of irradiated fuel which is

more economic compared with the nitric acid processing route.

It is a further object of the present invention to provide

a method for the reprocessing of irradiated fuel which does

not have the problems of corrosion and toxicity of earlier molten halide salt methods.

According to the present invention, there is provided a method for the processing of oxide nuclear fuel containing uranium, the method including the step of reacting the oxide with a fused alkali metal carbonate to produce a

compound which may be further processed so as to extract at

least a part of the uranium therefrom.

The carbonate may comprise sodium carbonate, Na,C0 ₃ ;

potassium carbonate, K*,C0 ₃ lithium carbonate, Li ₂ C0 ₃ ; or

any suitable mixture of these or other salts such as, for

example, an eutectic mixture of sodium and potassium

carbonates . Mixtures may be chosen on the basis of their

particular chemical reactivity in a particular environment

and/or on the basis of the melting temperature or melting

temperature range relative to a further intended processing

step.

In order to assist in the reaction or dissolution of the

oxide fuel material, additional chemicals may be added.

Such chemicals may include an oxidising agent such as

potassium nitrate, sodium peroxide or potassium oxide for

example.

Alternatively or in addition, air or oxygen may be passed

through the molten salt in order to assist in the reaction

of the fuel with the carbonate.

The molten carbonate salt may also contain sodium hydroxide or other alkali compound as appropriate in order to assist

dissolution of the fuel oxide.

The molten carbonate salt may also include a proportion of alkali metal halide such as sodium chloride or potassium

fluoride, for example, in order to assist dissolution of the oxide fuel into the carbonate.

Dissolution of the irradiated fuel material may be enhanced by a further processing step of initially oxidising the fuel oxide to convert it into uranium oxide powder of composition U ₃ 0 ₃ or U0 ₃ by known techniques.

Once dissolution of the oxide fuel into the carbonate has been achieved, the mixture may be further processed so as to remove the uranium, plutonium and neptunium, if present, from the mixture. Such further processing may normally

involve the electrolysis of the fused salt mixture, such electrolysis being in two stages.

A first electrolysis stage may be to remove at least a

proportion of the uranium present in the fused salt mixture

by deposition of either uranium metal or uranium oxide,

DO*,, at the cathode. The effect of this first electrolysis stage will be to remove at least some of the uranium from the fused salt mixture and so to increase the proportion of

plutonium therein. Irradiated thermal reactor fuel normally has in the region of 1% plutonium therein whereas new fuel may have about 6% plutonium or more therein. Therefore, electrolysis may preferably be continued so as to increase the relative proportion of plutonium to uranium to substantially above 6%.

A second electrolysis stage may employ a different cathode so as to allow the extraction of at least the plutonium, the plutonium and most of the remaining uranium migrating thereto and alloying therewith. Such a cathode may comprise, for example, molten cadmium metal. At the end of the second electrolysis stage, the resulting cadmium/ plutonium/ uranium (/neptunium) cathode is removed and the unwanted cadmium removed, for example, by distilling the cadmium off. The remaining fuel metal elements being further processed into the desired final fuel oxide

material by known techniques.

After the electrolysis stages, most of the remaining fused salt will be recycled to react with further portions of irradiated fuel. In order to prevent fission products from accumulating to unacceptable levels, a portion of the fused

salt will be discarded as waste at each cycle, and with it

the fuel cladding and other unreacted debris.

An advantage of the method of the present invention is that

the wastes produced are less mobile in a disposal environment than the alkali metal halides of the prior art and are thus easier to dispose of.

The method of the present invention may be operated as a batch process, the process employing a first step of making a required fused carbonate salt mixture and adding the irradiated fuel thereto in the required proportion to achieve at least partial dissolution of the fuel oxide therein; a second step may be the first electrolysis stage

as described above,- a third step may be the second electrolysis stage as described above,- and, a fourth step may include the further processing of the resulting uranium and/or plutonium and any other metal by-products into the

final desired products such as U0 ₂ powder or PuO, powder for the manufacture of new fuel rods for example.

The first and second electrolysis stages may be carried out

in a common electrolysis cell having different cathodes for

each stage, or separate electrolysis cells may be employed

for each stage.

The carbonate salt may be melted and maintained molten by

external heating means on the electrolysis cell or may be

melted by resistance heating means in the salt.

The molten cadmium cathode may be maintained molten by the

conduction of heat from the fused salt or by separate

heating means applied thereto.

An alternative to the electrolysis stages described above

may be the displacement of uranium and plutonium from the

fused salt melt by the addition of a refractory acidic

oxide. Such oxides may include silica and phosphorus

pentoxide, for example.

Such a method of extraction may be described by the

expression:

Na ₂ U,0 _{7 ( 1} *. + SiO _: Na,Si0 _{3 i l} , + 2U0 ₃

In order that the present invention may be more fully understood, an example, will now be described by way of illustration only with reference to the accompanying drawings, of which:

Figure 1 shows a flow diagram of an embodiment of a

reprocessing cycle according to the method of the present invention;

Figure 2 shows a schematic cross section of an electrolysis cell for carrying out a first electrolysis stage,- and

Figure 3 which shows a schematic cross section of an electrolysis cell for carrying out a second electrolysis stage.

Referring now to the drawings and where Figure 1 shows a flow diagram indicating the steps involved for an

embodiment of the method according to the present

invention. Irradiated fuel rods are first chopped into

short lengths or otherwise breached (10) . A carbonate

mixture of the eutectic composition of sodium and potassium

carbonates is made up with any required additional

chemicals to assist dissolution of the fuel (12) . The

breached fuel rods are added to the fused carbonate mixture

so as to react therewith and to achieve at least partial

dissolution therein with or without further additions of

supplementary chemicals (14) . Optionally some of the fused

carbonate from a previous cycle may be added (32) . Once at

least partial dissolution of the oxide fuel has been

achieved, the first electrolysis stage is carried out (16)

which produces solid uranium or uranium oxide on the cell

cathode (18) , this stage being described in more detail

with reference to Figure 2 below. The solid uranium or

oxide is then further processed as desired according to

known techniques. The remaining fused carbonate salt

mixture from 16 is then subjected to a second electrolysis

stage (20) , this stage being described in more detail with

reference to Figure 3 below. The second electrolysis stage

employs a molten metal cathode such as cadmium which

results in plutonium, uranium and possibly neptunium

alloying therewith so as to extract the bulk of the remaining metal from the fused salt (22) . The remaining fused salt is then treated to separate the debris of the

fuel rod cans and some of the carbonate from the bulk of the carbonate (28) , the debris and some carbonate waste being discarded (30) and the remainder of the carbonate

being recycled together with new carbonate make-up (32) .

The extracted plutonium/ uranium/ neptunium/ cadmium alloy from 22 is then processed by heating and distilling off the cadmium (24) to leave the required product (26) .

Figure 2 shows a schematic cross section through an

electrolysis cell for carrying out a first electrolysis stage. The cell comprises a vessel 40 having therein a

fused carbonate salt mixture 42 in which an irradiated oxide fuel has been reacted so as to achieve at least partial dissolution. An anode 44 and cathode 46 are provided together with a power supply and control apparatus 48. Heaters 50 are provided to melt and maintain molten the salt/oxide fuel mixture 42. Electrolysis of the mixture 42 results in uranium or uranium oxide 52 being deposited on

the cathode 46 so as to deplete the uranium content in the mixture 42. If all of the oxide fuel was not initially

reacted or dissolved in the fused carbonate salt, as the uranium content drops during electrolysis, more of the oxide fuel will react with the salt mixture or reacted but

undissolved material will progressively dissolve.

The initial reaction of the oxide fuel with the carbonate and oxygen, either from an oxide in the fused salt mixture or from air or oxygen passed through the molten salt may be typically described by the reaction:

2UO _:lβl + 0, _(g , + Na,C0 ₃₍₁₁ *■ Na ₂ U ₂ 0 ₇₍₁ , + C0 _2(g)

In the first electrolysis stage to extract uranium, the reaction may be described by:

U ₂ 0 ₇ ^2' + 4e ^~ ► 2U0 _2(S) + 30 ^2" ; or

U _: 0 ₇ ^2" + 12e ^" ► 2U _(s) + 70 ^:" ; depending

upon whether uranium oxide or metallic uranium is deposited

on the cathode.

1 ?

Figure 3 shows a schematic cross section through an electrolysis cell for carrying out a second electrolysis stage of the method according to the present invention. The

cell comprises a vessel 70 having heaters 72 to maintain molten a salt mixture 74 resulting from the first electrolysis stage as described above with reference to

Figure 2 , and also to maintain molten a cadmium cathode 76.

An anode 78 and power supply and control apparatus 80 are provided. During electrolysis, plutonium ions and uranium ions (and neptunium ions if present) migrate to the molten cadmium cathode and alloy therewith. After all the metal which can be extracted has been, the cadmium cathode is removed and the cadmium metal removed by distilling off at high temperature to leave behind the required metals which may be further processed according to known techniques to produce new fuel rods.

The reaction to extract plutonium from the fused salt during the second electrolysis stage may be of the form:

Pu,0 ₇ ^*" - 12Cd ₍₁ , + 12 e ^" ► 2PuCd ₅₍₁ , ÷ 70"