Partitioning of the spent nuclear fuel

reactor, the uranium in the fresh fuel feed does not require any enrichment. At the same time, very high burnup levels can be achieved through the core arrangements with dedicated regions where Pu can be effectively bred and the regions where Pu generated in former region can be effectively fissioned [Ficher G.J. et al., 1979]. However, B&B concept employs fast spectrum to provide effective breeding of Pu. As a result, Pu in the spent fuel is rich in Pu239 isotope (over 80% of total Pu [Yarsky P., 2003]) which raises significant proliferation concerns.

As discussed above, selected issues of the once through fuel cycle can be addressed via extended fuel burnup. However, the extent of these benefits is limited by the fuel performance with respect to its ability to withstand radiation damage and exposure to severe reactor operating conditions for long periods of time. Alternative fuel designs such as coated TRISO particles suggested for use in advanced gas cooled reactors (Pebble Bed Modular Reactor – PBMR [Koster A. et al., 2003] and Gas Turbine-Modular Helium Reactor - GT-MHR [LaBar M. P., 2002]) can maximize the benefits of the high burnup in the once through fuel cycle.

Partitioning of the spent nuclear fuel

Recycling of the SNF is the ultimate approach to diminishing the concerns over the nuclear waste in the once through fuel cycle to a considerable extent. Even after irradiation, the major part of the fuel is uranium (Figure 1.2.1). Uranium is practically non radioactive and if separated from the rest of the irradiated fuel constituents with adequate efficiency can be stored as a low level waste (LLW) or reintroduced into the fuel cycle. This would allow a major reduction in volume of the SNF and enhance the effective storage capacity of geological repository.

Pu constitutes about 1% of the SNF. It can be used as fuel virtually in any type of reactor. Recycling plutonium in LWRs is a common practice in Europe, Russia, and Japan. Other countries also consider Pu recovery and develop SNF reprocessing technologies and infrastructure for that purpose. Typically, recovered Pu is mixed with depleted or natural uranium in UO2-PuO2 mixed oxide (MOX) form. Currently, the spent MOX fuel assemblies are not reprocessed. Nevertheless, even single path Pu recycling results in significant improvement in natural uranium utilization and proliferation resistance characteristics [Pellaud B., 2002].

Fission products typically amount to less than 4% of the spent fuel but most of them are stable or very short lived. Only about 0.4% of the fission products are of a significant importance (Figure 1.2.1). Relatively short lived Cs137 and Sr90 are responsible for most of the decay heat production in the first few decades after fuel discharge from the reactor. Separation of Cs and Sr from the spent fuel constituents for dedicated storage would practically eliminate the heat load management issue in the repository design. The activity of all fission products collectively decreases below the level of natural uranium ore required to manufacture the initial fuel in a few hundreds of years (Figure 1.2.2).

95.6% is uranium – can be dispositioned as Class C low-level waste or recycled

3% is stable or short lived fission products that do not pose major disposal challenges

0.3% is cesium and strontium that decay out in a few centuries (and are the primary near term HLW heat source)

0.1% is long lived iodine and technetium which can be transmuted in thermal reactors

0.9% is plutonium which can be burned as a fuel in any reactor

0.1% is long lived actinides which can be effectively fissioned in fast spectrum reactors or accelerator driven systems (ADS)

Figure 1.2.1. Typical composition of the spent nuclear fuel [DOE/NE, 2003].

Long lived fission products (LLFP) I129 and Tc99 represent a small fraction of the SNF (about 0.1%). These isotopes were identified as significant contributors to the long term radiation dose from the repository due to their high solubility in water [Van Tuyle G.J., 2001]. However, Tc99 and I129 can be transmuted to short lived isotopes if subjected to epithermal neutron flux. A number of studies showed the possibility of Tc99 and I129 transmutation in both thermal and fast spectrum reactors [Brusselaers P. et al., 1996], [Krivitsky I. and Kochetkov A. L., 2000], [Hejzlar P. et al., 2001]. The consensus on whether fast or thermal spectrum systems are more efficient with regards to LLFP transmutation has not yet been reached and different LLFP transmutation concepts are still under investigation. The factors affecting the best concept choice are: transmutation rates, achievable burnup fractions, available technologies, and the effects of LLFP

transmutation targets on other performance characteristics of transmutation system. A transmutation system of any type will require increased fissile inventory or external neutron source to compensate for the parasitic neutron absorption in the targets containing LLFP.

Radiotoxcicity Relative to NU

1E+06

1E+05

1E+04

1E+03

1E+02

1E+01

1E+00

1E-01

Nuclear Waste (FP only) Nuclear Waste (FP+Actinides)

UO2 Fuel Enrichment =4.2% (0.3% waste assay) Burnup=50 MWd/kg

1E-02

1E-03 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06

Decay Time, Years

Figure 1.2.2. SNF Radiotoxicity for ingestion as function of time after discharge.

Only about 0.1% of the spent fuel is minor actinides (MA). Some of them are very long lived. In the repository, Pu and MA (also regarded as Transuranic elements – TRU) are responsible for most of the radiotoxicity of the SNF after decay of the fission products in the period between 1000 and 1M years after discharge (Figure 1.2.2). In the countries practicing fuel reprocessing, the high level radioactive waste, including MA and all the fission products, is incorporated in chemically stable host matrix (typically vitrified in glass). Only such vitrified waste is intended for geological storage. This fuel cycle strategy alone allows - major reduction in volume of high level nuclear waste (by about 40% in France) - improvement in uranium utilization through single path Pu recycling as MOX fuel - enhancement of proliferation resistance characteristics of the fuel cycle due to degradation of Pu isotopic vector [Pellaud B., 2002]