Didier Haas, M. Hugon, and M. Verwerft

Abstract Since the early 1970s, studies and experimental projects have been undertaken in Europe to examine the potential of thorium-based fuels in a variety of reactor types. The first trials were mainly devoted to the use of thorium in high — temperature reactors. These projects can be seen as scientific successes but were not pursued on a commercial basis because of the priority given in Europe to the development of light water reactors. Later on, thorium oxide was considered as a potential matrix for burning plutonium (possibly also minor actinides), and several core design studies, as well as experiments, were undertaken. The most recent such concern the BR2 and HFR Material Test Reactor (MTR) irradiations in Belgium and in the Netherlands, respectively, as well as the KWO PWR in Obrigheim in Germany, in which thorium-plutonium oxide fuel (Th-MOX) was successfully irradiated up to 38 GWd/tHM. The results of these experiments have shown that Th-MOX behaves in a comparable way as conventional uranium-plutonium oxide fuel (U-MOX). More work is still needed before Th-MOX will reach sufficient maturity to implement it on a large scale in power reactors, but all currently available results indicate that licensing Th-MOX for LWRs should be feasible. Finally, European research projects are still devoted to the study of thorium salts in molten salt reactors, a design that incorporates on-line reprocessing and needs no specific thorium fabrication, adding therefore the benefits of thorium without its main challenges.

Keywords Fuel • HTR • LWR • MSR • Plutonium • Thorium

International Symposium on Nuclear Back-end Issues and the Role of Nuclear Transmutation Technology after the accident of TEPCO’s Fukushima Daiichi Nuclear Power Stations

D. Haas (*)

Consultant, 101 rue de la Station, 1457 Walhain, Belgium e-mail: Didier. haas@hotmail. be

M. Hugon

European Commission, Brussels, 1049 Brussels, Belgium M. Verwerft

SCK*CEN, Boeretang 200 B-2400 Mol, Belgium © The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_18

18.1 Introduction

Natural thorium (Th) has only one isotope, 232Th, which is fertile. In a thermal reactor, Th can absorb neutrons and, following nuclear reactions, produces 233U, which is fissile. Under optimized breeding conditions, a sustainable Th-233U cycle can be reached, but the thorium cycle needs a seed or driver fuel, which can be based on 235U or on Pu.

233U 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 U rather than U or 9Pu, 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 233U 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 232U 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.

This presentation summarizes the history and status of the main European research programs (cordis. europa. eu) with Th use. These programs concerned HTRs, LWRs, and MSRs. Emphasis is given here on the latest two developments.