18.1 Introduction
Natural thorium (Th) has only one isotope,
232 Th, which is fertile. In a thermal
reactor, Th can absorb neutrons and, following nuclear reactions, produces
233
U,
which is fissile. Under optimized breeding conditions, a sustainable Th233 U cycle
can be reached, but the thorium cycle needs a seed or driver fuel, which can be
based on
235 U or on Pu.
233 U as a fissile nuclide features high neutron production in a thermal and
epithermal neutron spectrum. This ability offers improved neutron economy for
reactors fueled with
233 U rather than
235 U or
239 Pu, particularly at thermal energies
in light water reactors (LWRs). In theory, breeding (formation of fissile nuclides) is
achievable at thermal energies with a Th/233U fuel, which is not the case with
U-MOX fuel. However, even though breeding can be demonstrated at an experimental level, optimal breeding is not achieved in the current fleet of LWRs. In
today’s context, U-MOX fuels are not reprocessed, and here Th-MOX offers
perhaps its best advantage over U-MOX. The excellent chemical stability of the
thorium oxide matrix makes it an excellent candidate for direct disposal, and thus
also for once-through fuels allowing burning excess Pu without production of
higher actinides. Another alternative would be to use Th-MOX fuels in LWRs as
a means to initiate the breeding of
233 U for future use in other reactor types, as an
option to save natural U and to further improve the U-Pu fuel cycle.
In addition to the LWR/FR scenario, two reactor types have been considered for
a breeding Th fuel cycle in the future: high-temperature reactors (HTRs) and
molten salt reactors (MSRs). HTRs represent the fastest route to implement a closed
breeding Th fuel cycle. The technology exists conceptually but needs to be developed before commercialization (which is pending). Also, supporting technologies
associated with fuel manufacturing, reprocessing, transport, waste management,
and final disposal need to be developed. MSRs represent a longer-term development option for Th fuel cycles. In MSRs loaded with Th-based fuels, breeding may
be achieved over a wide range of neutron energies. On-line reprocessing is an
important feature of MSRs, which enables continuous re-use of the nuclear fuel by
extracting the fission products.
The potential development of a closed Th fuel cycle faces some obstacles.
Reprocessing is one of these, as Th oxide is more stable than U oxide. In contrast
to the Purex process, which has been industrially operational in the U-Pu fuel cycle
for more than 30 years, the Thorex process, which has been investigated for many
years in laboratories, faces some difficulties: it requires stronger acids (and therefore more advanced corrosion-free materials for process vessels) and longer dissolution times. Remote-controlled fuel manufacturing represents another challenge as
Th-based fuels have high-energy gamma radiation from the presence of
232 U after
irradiation, which requires remote fabrication and handling in heavily shielded
facilities. Thus, this fuel fabrication, transport, and reprocessing are more complex
than the present practice for U oxide fuel, for instance.
198
D. Haas et al.
Natural thorium (Th) has only one isotope,
232 Th, which is fertile. In a thermal
reactor, Th can absorb neutrons and, following nuclear reactions, produces
233
U,
which is fissile. Under optimized breeding conditions, a sustainable Th233 U cycle
can be reached, but the thorium cycle needs a seed or driver fuel, which can be
based on
235 U or on Pu.
233 U as a fissile nuclide features high neutron production in a thermal and
epithermal neutron spectrum. This ability offers improved neutron economy for
reactors fueled with
233 U rather than
235 U or
239 Pu, particularly at thermal energies
in light water reactors (LWRs). In theory, breeding (formation of fissile nuclides) is
achievable at thermal energies with a Th/233U fuel, which is not the case with
U-MOX fuel. However, even though breeding can be demonstrated at an experimental level, optimal breeding is not achieved in the current fleet of LWRs. In
today’s context, U-MOX fuels are not reprocessed, and here Th-MOX offers
perhaps its best advantage over U-MOX. The excellent chemical stability of the
thorium oxide matrix makes it an excellent candidate for direct disposal, and thus
also for once-through fuels allowing burning excess Pu without production of
higher actinides. Another alternative would be to use Th-MOX fuels in LWRs as
a means to initiate the breeding of
233 U for future use in other reactor types, as an
option to save natural U and to further improve the U-Pu fuel cycle.
In addition to the LWR/FR scenario, two reactor types have been considered for
a breeding Th fuel cycle in the future: high-temperature reactors (HTRs) and
molten salt reactors (MSRs). HTRs represent the fastest route to implement a closed
breeding Th fuel cycle. The technology exists conceptually but needs to be developed before commercialization (which is pending). Also, supporting technologies
associated with fuel manufacturing, reprocessing, transport, waste management,
and final disposal need to be developed. MSRs represent a longer-term development option for Th fuel cycles. In MSRs loaded with Th-based fuels, breeding may
be achieved over a wide range of neutron energies. On-line reprocessing is an
important feature of MSRs, which enables continuous re-use of the nuclear fuel by
extracting the fission products.
The potential development of a closed Th fuel cycle faces some obstacles.
Reprocessing is one of these, as Th oxide is more stable than U oxide. In contrast
to the Purex process, which has been industrially operational in the U-Pu fuel cycle
for more than 30 years, the Thorex process, which has been investigated for many
years in laboratories, faces some difficulties: it requires stronger acids (and therefore more advanced corrosion-free materials for process vessels) and longer dissolution times. Remote-controlled fuel manufacturing represents another challenge as
Th-based fuels have high-energy gamma radiation from the presence of
232 U after
irradiation, which requires remote fabrication and handling in heavily shielded
facilities. Thus, this fuel fabrication, transport, and reprocessing are more complex
than the present practice for U oxide fuel, for instance.
198
D. Haas et al.