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amounts of MAs. In addition, one should keep in mind that, as mentioned above, the very high energy
part of the spallation neutron spectrum was not taken into account, neither in Reference [15] nor in
the present study.
6. Required transmutation capacity and associated fuel cycle facilities [16]
A 100 GWe nuclear LWR UO 2 reactor park produces annually 2 200 t spent fuel. This inventory
contains approximately 25 t Pu, 1.6 t Np, and 1.6 t Am+Cm. If nuclear electricity production
continues during an indefinite time period, the spent fuel inventory will continuously grow and
become a large nuclear legacy which will have to be properly managed during a very long period,
exceeding human civilisation. To decrease the growth rate of this spent fuel inventory, particularly its
incorporated Pu mass, reprocessing of LWR-UO 2 is a first step to reduce the accumulation rate.
However the HLW resulting from reprocessing still contains the 3.2 t of MAs. These nuclides
constitute the long-term radiotoxic inventory of the high level waste (HLW) if no partitioning is
performed.
Partitioning of the MAs, followed by transmutation-incineration, could achieve a 10 to 100-fold
reduction of the residual radiotoxicity. The recycling of separated MAs can be envisaged by mixing
Np with the LWR-MOX fuel or, more effectively, by mixing all the MAs with FR-MOX fuel. A fuel
fabrication capacity of 60 t FR-MOX-(2.5%Np) and 60 t FR-MOX-(2.5%Am) would have to be
installed near the reprocessing plants. A composite reactor park containing 70 GWe-LWR-UO 2 ,
10 GWe-LWR-MOX and 20 GWe-FR+ADS is capable of stabilising the TRU inventory. Reduction
of the MA fuel mass to be handled in the FR-ADS systems is still possible if higher concentrations of
MAs can be introduced into the sub-critical cores of ADS systems. An 820 MW th ADS core (1.5 GeV,
40 mA) containing 60% MA and 40% Pu could transmute 250 kg TRU per year. With a thermal to
electric yield of 30% it would produce 246 MWe which would in part be recycled to the accelerator
(146 MWe) and in part delivered to the grid. Gradually the ADS capacity should increase from
8 GWe initially to 20 GWe at the end of the nuclear energy production to cope with the entire residual
TRU inventory.
7. Conclusions
With a routine thermal output of 60 MW th , BR2 has a limited irradiation potential for 12 targets
of 500 g Np+Am each and a transmutation throughput of the order of 1.5 kg Np+Am per 200 EFPD.
This transmutation capacity can be used for investigating, at the technological scale, the formation of
transmutation products ( 238 Pu, 239 Pu, FPs...) in a thermal neutron spectrum with large contribution of
epithermal and fast neutrons as well as the metallurgical behaviour of the targets. In particular, if the
irradiations are carried out during a long period, the calculated high fission-over-total-disappearance
rate in the 237 Np-and Am targets could be checked. It is indeed essential to take into account the total
length of the transmutation chains when performing calculations for high flux reactors.
One of the purposes of MYRRHA is its utilisation for the investigation of actinide transmutation
feasibility with ADSs. With a total power not to exceed 30 to 35 MW, fast fluxes (E>0.75 MeV) up to
10 15 n/cm 2 s are to be attained in irradiation positions near the spallation source. Calculations indicate
lower transmutation rates in MYRRHA than in BR2, but fast spectrum systems, and in particular ADS
devices, are characterised by better neutron economics in the transmutation process, i.e. by a higher
“direct” fission-over-total-disappearance rate. MYRRHA as multipurpose ADS for R&D is hence an
interesting tool to investigate transmutation of MAs in a fast neutron environment, with the
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