BUC applications for MOX

3.4.3.1 Transportation

For MOX fuel, the problem of transportation is mainly that of shielding, rather than that of criticality issue. This is why during transport it is necessary to mix MOX with UOX fuel.

Even with this difficulty, BUC can be used for the design of new transport flasks.

3.4.3.3 Disposal

For disposal of MOX fuel, there is no problem of shielding, so it initially appears that one may be able to take greater advantage of BUC than in transport operations. However, over the longer term (greater than 100 years), the Pu-240 and Am-241 content of MOX fuel decreases, which leads to an increase in the reactivity.

Burnup k-infinity

UOX / EPR MOX

WG-MOX RG-MOX

Table VI. Fuel Cycle and Disposal Path for Fast Reactors Worldwide

Location Reactor Fuel type Fuel cycle Disposal Path

USA FFTF MOX Once through Yucca Mountain

France Phenix MOX Reprocess Recycle in LWR

France Super Phenix MOX Reprocess Recycle in LWR

Russia BN 350 UO2 Once through

Long term storage

Possible reprocessing at Mayak

Russia BN 600 UO2 conversion to

MOX Once through

Long term storage

Possible reprocessing at Mayak

Japan MONJU MOX Reprocess Recycle in LWR

USA EBR II U and Pu metal Treatment Waste forms to geological repository

This is a greater problem for MOX fuel than for UOX fuel as the latter does not contain as much Pu-240. For UOX fuel, the reactivity does not change significantly with time. However for MOX fuel, the reactivity can be greater after 50,000 years than it was initially (Fig. 2). Due to the slightly different fissile isotopes associated with MOX, than UO2 based systems direct disposal of MOX fuels will force an additional look at the near-field geochemical development. The US has investigated direct disposal of MOX and has concluded that direct disposal for the proposed US MOX design is feasible at the proposed Yucca Mountain site.

3.4.3.3 Storage

BUC may be more easily applied to the storage of MOX fuel in the medium term due to the reduction in reactivity of the fuel over the initial 100 years period. A distinction cannot be made between wet and dry storage. In many countries, dry storage containers must comply with the IAEA recommendations (TS-R-1) [7] for transport. These recommend that the presence of water is considered within the container. For reactor-grade MOX, storage facilities and shipping cask designed for UO2 at the criticality safety point of view does generally accommodate MOX fuel without any problem. For weapons grade (WG) MOX and/or over moderated MOX fuel assemblies, BUC could be advantageously considered.

3.4.3.4 Reprocessing

As for UOX fuel, the advantage of BUC is to avoid or minimize the need of gadolinium in order to maintain the reprocessing rate. Since the residual fissile content of spent MOX fuel is higher than for UOX fuel, BUC considerations are essential to ensure the capacity and performance of the reprocessing operations.

Fig. 2. K-infinity vs. decay time for MOX and UO2 fuels.

3.4.3.5 Important isotopes associated with fast breeder MOX BUC

As actinide-only BUC is not a viable option for fast breeder MOX fuel, the presence of fission products needs to be considered. The most important fission product is Sm-149, which is produced, in large quantities, and contrary to LWR fuel, the thermal capture rate of Sm-149 is negligible during irradiation.

After irradiation, in applications where water is present or needs to be considered, the thermal absorption of Sm-149 will be particularly effective. Consequently the use of ‘samarium’ BUC will be relevant.

3.4.3.6 Additional research, developments and operational needs for MOX

Research and development concerning BUC of MOX fuel should focus on the following topics:

• Depending on the application of BUC, the calculation methodologies for determination of the spent fuel inventory for MOX and UO2 may require additional study:

- Well assessed methods and data sets, concerning particularly shielding of actinides, cross-section sets, nuclide chains,

- Geometrical modelling taking into account the environment of the fuel during irradiation;

• Improvement of spent fuel isotopic composition database. Compared to UO2 fuel, for which a wide range of PIE experiments have been performed for years, the MOX databases should be further investigated;

• Special attention should be paid to the minor actinides, especially americium and curium isotopes. The highly fissile minor actinides (Am-242m, Cm-243, Cm-245) could

Decay time (year, log scale) k-infinity

10² 10⁵ 10⁷

UO2 MOX

contribute to several percent to the reactivity at the end of life, especially for high burnup and high Pu-content MOX fuel. Present comparisons between calculation and measurement show large discrepancies (10-20 %) on the inventory of these isotopes;

• If BUC measurements are required, burnup measurement devices for MOX may need to be enhanced in order to be comparable to those used for UO2 fuel;

• Concerning final disposal of spent MOX fuel, the geochemical behaviour of plutonium and minor actinides must be studied, as well as the long term reactivity evolution of the fuel. Studies performed in the US indicate that little or no impact to the performance of the repository will occur due to direct disposal of MOX fuel [8].