Differences between the effects of high burnup UOX and MOX on spent fuel management

5.2.1. Spent MOX storage

Due to the mechanical designs of MOX elements being identical to those of UOX fuel, MOX may be stored in the same facilities and often alongside UOX assemblies. The primary challenges from spent MOX relative to

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FIG. 40. Radioactivity (solid line) and α-radioactivity (broken line) of SF of UOX, MOX and ERU at five years after SF discharge [24].

UOX fuels are decay heat and neutron activity, due to minor actinide content, as shown in Figs 41 and 42. The increase in ²⁴⁴Cm content requires additional neutron shielding, while the increased decay heat must be removed by the cooling systems.

Dry storage of spent MOX may be more challenging. As noted in Section 4.2.3, the end of life pressure in MOX fuels tends to be higher than that of UOX of equivalent burnup. In combination with increased decay heat, the driving force for cladding creep is potentially higher than that for HBU UOX. This may be compounded by a need to add additional neutron shielding to the storage facility.

Two strategies may be considered to mitigate this problem. Increasing the pond storage time prior to a move to dry storage may allow the decay heat to drop to acceptable levels, however, as shown in Fig. 41, the rate of decrease is relatively slow and therefore, this strategy may not release sufficient pond space in time. In addition, MOX elements may be mixed with UOX fuel in the same storage facility or cask to balance decay heat and neutron generation, with fuel assemblies being used as part of the shielding design.

To date, it is generally more convenient for utilities to load storage casks with lower burnup UOX fuels and retain MOX assemblies in the pond, however, some MOX elements have been loaded to casks and there is little evidence of difficulties with this.

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FIG. 41. Decay heat of SF of UOX, MOX and ERU at five years after SF discharge [24].

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FIG. 42. Neutron emission of SF of UOX, MOX and ERU at five years after SF discharge [24].

5.2.2. MOX fuel reprocessing

Section 5.1.3 presented the impact of higher burnup UOX fuels on reprocessing plants. Many of the issues discussed are also applicable when reprocessing MOX fuels, modified for the differing actinide and fission product contents. The discussion below presents the effect of spent MOX fuel reprocessing on the factors considered in fuel handling prior to dissolution and heat removal.

MOX elements have the same mechanical design as UOX fuel and present no new physical handling problems. However, as shown in Fig. 41, the decay heat of MOX fuel decreases more slowly than that of UOX fuel due to increased minor actinide content. Therefore, MOX fuels may need to be cooled for longer prior to reprocessing, or more likely, blended with UOX fuels to allow the heat loads to be managed within the plant.

Radiometric instruments will need to be recalibrated when MOX fuel is fed to the reprocessing plant, however, in this regard, the increased passive neutron emissions of MOX fuel may be used as a differentiator from UOX fuel and used to control the feed order and blending strategy.

5.2.2.1. Dissolution

There are two key differences in the dissolution behaviour of MOX fuels compared with UOX. MOX fuels generate higher levels of insoluble fission products due to Pu fissions and differences in fuel rod chemistry [74, 79].

As discussed in Sections 3.1.2 and 4.2.3, the manufacturing processes used to fabricate MOX fuels result in Pu-rich islands within the pellet structure. While these islands are depleted by irradiation they retain a high Pu concentration which may make them difficult to dissolve [83]. Many modern MOX fabrication methods limit the local Pu concentration to ~30 wt %, the point at which PuO₂ dissolution becomes more challenging [84]. On dissolution any undissolved Pu rich particles are released into the product liquor and it is important to demonstrate that over a period of time PuO₂ particles do not accumulate within process vessels to produce a criticality hazard.

The size range of both the insoluble fission products and the PuO₂ particles mean they are generally captured by the centrifuge step within the reprocessing plant. Therefore, the centrifuge cake from MOX fuel will contain both increased IFPs and Insoluble Pu solids.

5.2.2.2. Solvent extraction

As shown in Fig. 40, the α activity of MOX fuel is much higher than that of UOX fuel of equivalent burnup.

Therefore, the energy passed to the solvent, if processing MOX fuels is significantly increased. The higher plutonium concentrations present when reprocessing MOX fuel may also lead to increased rates of parasitic and autocatalytic reactions and increase the risk of plutonium recycle within contactors and the accumulation of unacceptably high plutonium concentrations in particular stages. Increased feeds of reducing agent to the uranium-plutonium separation contactor are required in order to accommodate the higher uranium-plutonium concentration and increased levels of stabilizing agents; e.g. hydrazine, may be required to prevent oxidation of species such as U(IV) and Pu(III).

Again, a relatively simple way of overcoming issues related to the reprocessing of MOX fuel in the current generation of reprocessing plants is to adopt a blending strategy in which MOX fuel is mixed with a larger amount, perhaps three or four times as much, uranium dioxide fuel in the early stages of reprocessing, prior to solvent extraction. Alternatively, different solvent contactors with faster throughput rates that minimize residence time may be considered.

5.2.2.3. Product recovery

The process chemistry of product recovery is essentially unchanged when considering the liquors derived from MOX fuels or blended feeds. The key limit tends to be the throughput of the Pu powder production plant.

Typical industrial reprocessing plants have designs suited to total daily throughputs of approximately 5 tons of UOX fuel and plutonium contents of 1% or 50 kg Pu per day. The amount of plutonium present in a typical spent MOX fuels is typically several weight percent and therefore redesign of plutonium recovery operations (precipitation, drying, calcining and canning) may be necessary to accommodate the increased amounts of plutonium present.

As discussed in Section 5.1.6, the α activity of separated Pu derived from MOX reprocessing is increased relative to that of standard UO₂ and similar considerations regarding radiolysis and heat during storage apply.

5.2.2.4. Waste treatment

As with UOX fuel, the amount of fission product waste in MOX increases linearly with burnup, however, the amount of Pu and minor actinides is higher. Within the HLW area, the main challenge is ²⁴⁴Cm and neutron activity, leading to potential reductions in waste incorporation rates in glass containers if the liquors cannot be blended with less onerous material [85]. For ILW, the main challenge is in the centrifuge cake stream as noted above due to increases in IFPs and the presence of PuO₂ rich solids [83].

As with UOX fuels, providing process decontamination factors (DFs) remain the same, the amounts of active species released to the environment generally depend on the fuel burnup, however, the precise nature of the discharge will be modified to reflect the fission products produced by Pu fissions. For instance, aerial discharges of

85Kr may be reduced, while ¹²⁹I releases could increase. The main challenge to acceptable discharges if reprocessing MOX fuel is maintaining efficient operation in the solvent extraction processes as previously discussed.

Many of the management strategies described above depend on the blending of MOX with UOX fuels to reduce the overall activity (particularly α and neutrons) of the process liquors. While the amount of low burnup UOX in fuel ponds is reducing over time, UOX blend stock will continue to be available while the practice of 1/3 MOX core loadings is maintained in PWR reactors.

As with a move to higher burnup UOX fuel, spent MOX is more challenging to current reprocessing plants, per ton of fuel processed. However, by recycling the Pu from spent UO₂ for irradiation, the need for further enrichment activities is reduced and Pu that would have been committed to a repository is reduced to shorter lived fission products. Therefore, as with high burnup UOX it is important to consider the advantages and disadvantages of MOX fuel utilization within the overall context of the fuel cycle and total energy generation rather than a narrow focus of management of current spent fuel.

Industrial MOX fuel reprocessing activities have recently started in France [86]. Around 10 t HM of MOX fuel irradiated to ~35 GW·d/t U were reprocessed in the UP2-800 plant. After modifications to the dissolution cycle, the fraction of insoluble Pu was limited to just over 0.01% of the Pu in the spent fuel. Prior to feeding to the extraction cycles, the MOX process liquor was diluted with uranium nitrate recycled from the uranium product stream in the plant. This blend was then processed through solvent extraction without incident.

5.2.3. Spent MOX transportation

The conditions of MOX spent fuel transportation would be very similar to ones described for high burnup UOX fuel transportation. Due to higher concentration of Pu and therefore higher emission of neutrons, there may be some changes in transportation conditions depending on criticality and neutron shielding calculations.

5.2.4. Spent MOX disposal

The impact of increased decay heat and ¹²⁹I production will have to be evaluated for MOX fuel disposal.

5.3. PROLIFERATION RESISTANCE

The proliferation attractiveness of high burnup UOX and MOX spent fuel from water reactor is clearly decreased. This is shown in the Chapter 4 of The Characteristics of Spent Fuel. The main formation chain of each plutonium isotope in the fuel is as follows:

92

Therefore, higher burnup in UOX fuels increases higher plutonium isotopes and already existing plutonium isotopes in MOX fuels accelerate higher plutonium isotopes further. The grades of plutonium material are classified into super; weapons, fuel, and reactor according to the content of the fertile isotope ²⁴⁰Pu are shown in Table 8.

The classification of the grade of plutonium is based on the high spontaneous fission neutron yield of ²⁴⁰Pu is shown in Table 9.

Fast neutrons released by the spontaneous fission of the doubly even isotopes ²⁴⁰Pu and ²⁴²Pu can initiate a premature chain reaction in the plutonium when the implosion of nuclear warhead material takes place. This is the

‘fizzle’ of the weapon, reducing its explosive yield [20].

The spent fuel of the enriched reprocessed uranium (REPU) contains more ²³⁷Np and hence more ²³⁸Pu than the enriched natural uranium due to the presence of ²³⁶U in the fresh fuel. Therefore it is more proliferation resistant.

It is felt that the degradation of the quality of recycled uranium could require additional protective measures because of the presence of the high energy gamma ray emitter ²³²U in the front end facilities where the material is treated before being reused in a reactor (enrichment facilities, manufacturing chains devoted to the use of REPU, transport containers). However no need for significant additional modification of such facilities is anticipated for the envisaged burnup extension frame. The situation could become little more critical in case of the multiple recycling of the same material [82]; however this is clearly not a short or medium term issue.