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Nuclear fuel reprocessing at the Sellafield nuclear reprocessing site

The objectives of nuclear fuel reprocessing are (a) to separate uranium from plutonium and (b) to separate a suite of highly active fission products i.e. gamma-emitters such as caesium-137 and cobalt-60 etc. from both uranium and plutonium. The reprocessing procedure followed by many reprocessing plants including BNFL is based on the Purex process and the initial stages include, (a) receipt bay and cooling, (b) decanning, (c) dissolution and off-gas treatment (d) primary separation, (e) uranium / plutonium separation and (f) plutonium purification (BNFL info. Sheet 1989). I will describe the reprocessing procedure for Magnox fuel rods only. The reprocessing of ceramic uranium oxide only differs in the initial dissolution stage due to the non-solubility of zirconium and stainless steel in a nitric acid medium (Jenkins et al 1984).

Firstly, irradiated fuel rods are transported to the receipt bay, discharged into rod skips and nominally stored under water for a minimum period of 150 days to allow for the decay of some of the shorter-lived nuclides such as iodine-131 and promethium-144 (Lambert 1990). The fuel rod is then stripped of its magnesium alloy sheath in the decanning plant and dissolved in nitric acid. The dissolution stage produces uranium oxide nitrate and plutonium nitrates which are either tetra or hexavalent. Dissolution reactions for both uranium and plutonium are shown below.

(1) 2UO2 + 4HNO3 +O2 >>>>>>>> 2UO2 (NO3)2 + 2H2O

 

U + HNO3 + 1.5 O2 >>>>>>>>>> UO2 (NO3)2 + H20

 

(2) PuO2 + 4HNO3 >>>>>>>>>>> Pu (HNO3)4 + 2 H2O

 

2PuO2 + 4HNO3 + O2 >>>>>>>>>>> 2PuO2 (NO3)2 + 2H2O

 

Some of the more refractory oxides of plutonium and the platinum group metals associated with the fission products may not fully dissolve during this stage and may cause low-efficiency separation problems in later separation and purification stages (Ballestra & Holme 1989). For ceramic uranium oxide fuel rods which are encased in stainless steel, considerable amounts of 'insolubles' are produced which are mostly associated with the noble metals fission group elements such as molybdenum, technetium and ruthenium (Jenkins 1983).

The solution containing dissolved plutonium, uranium and fission products then undergo 'clarification' to remove undissolved solids. Following clarification, Pu and U solutions are allowed to cool and then conditioned by adjustment of solution pH to ensure that Pu remains in its tetravalent state prior to complexion with tri-n-butyl phosphate (TBP) in odourless kerosene. The principal atmospheric wastes produced from these stages are the dissolver nitric gases which pass through two fume absorbers and a caustic scrubber (to capture iodine) before emission to atmosphere via the tall stacks of B204. Dissolver nitric acid liquid wastes are removed for disposal by vitrification. These wastes are defined as high level wastes (HLW).

As discussed earlier, the solvent used in the Purex process is TBP-kerosene which is able to complex hexavalent U and hexavalent and quadravalent Pu. By contacting the aqueous phase with an immiscible organic phase such as TBP, the unwanted metal ions (fission products) are extracted into the nitric acid medium whilst Pu and U move into the organic phase. The basis of the TBP-kerosene separation process is that fission products TBP do not complex with TBP, only Pu and U are able to do so.

The Purex separation process is shown below

(3) UO2+ + (a) 2NO3 + 2.2TBPo >>>>>>>>>>> UO2 (NO3)2.2TBPo

(4) Pu4+ + (a) 4NO3 + 2TBPo >>>>>>>>>>>> Pu (No3)4.2TBPo where (a) is aqueous phase and (o) is organic phase. Both reactions are reversible, the forward reaction only is shown.

The TBP solution is fed into a mixer and settling tank and the solution is mixed by a paddle and passed to a tank where the aqueous and organic phases separate via two strategically placed exit ports which are sited at different levels within the tank. Fission products associated with the aqueous phase are defined as 'raffinates' which are highly active. This waste is evaporated to reduce its overall volume prior to being vitrified or encased in a resilient glass-like holding medium.

The TBP phase containing U and PU are treated with ferrous sulphamate which reduces tetravalent Pu (4+) to its trivalent (3+) oxidation state. Meantime, U remains in its tetravalent state. Trivalent Pu is then stripped out of its organic phase and into an aqueous phase which effectively completes the Pu / U separation process. Following this stage, further cycles of back-wash and purification are carried out.

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