Th-MOX Fuels Irradiated in LWR Conditions

Within the European Framework Programmes, the study of Th fuels behavior in LWRs was first aimed at comparing the behavior and the applicability of various matrices to be used for the transmutation of Pu and minor actinides (projects THORIUM CYCLE, LWR-DEPUTY, OMICO). Comparisons were made with standard fuels (UO2, MOX), and also with so-called inert matrices fuels (using, for example, Mo or MgO as matrix in CERMET and CERCER fuel types, respectively). As explained earlier, irradiation experiments were performed in three facilities, namely, the KWO PWR, HFR, and BR2 Material Test Reactors.

The THORIUM CYCLE project was a 4-year project with the following participants: the coordinator NRG (NL), BNFL (UK), CEA (F), FZK and KWO (D), and JRC-IE and JRC-ITU (EU). The goals of this project, which started on 1 October 2000, were to supply key data for application of the Th cycle in LWRs. In particular, it included the study of• The behavior of Th-based fuel at extended burn-up through an irradiation experiment of four short fuel pins [UO2, (U,Pu)O2, ThO2, and (Th,Pu)O2] up to 55 GWd/tHM in HFR, and an irradiation experiment of one short fuel pin [(Th,Pu)O2] to 38 GWd/tHM in a PWR (KWO); it should be noted that a previous irradiation of (Th,Pu)O2 in Germany (Lingen) achieved a burn-up of 20 GWd/tHM [7];

• The core calculations for Th-based fuel, including code-to-code validation, sensitivity check for significant isotopes 232Th and 233U, and the calculation up to 80–100 GWd/tHM for Th-MOX fuel.

The irradiation test in KWO enabled the investigation of the operational safety of Th-MOX rod behavior under realistic pressurized water reactor (PWR) conditions. The short test rod was inserted in a MOX assembly to provide the most realistic boundary conditions possible. The foreseen MOX carrier assembly had already been irradiated for one cycle. The cladding appeared in good condition after irradiation, and its creep-down, measured at the reactor site during the shut-down periods, as well as its general behavior, were well within the bounds of experience for UO2 fuels. The fission gas (Xe and Kr) release was about 0.5 % [8], which is about half that for equivalent MOX fuels at the same burn-up, but the linear power was lower than in equivalent U-MOX studies. Taking into account experimental uncertainties, the fuel behavior seems to be at least as good as U-MOX.

The THORIUM CYCLE project was completed in 2006, but the postirradiation experiments were performed under a subsequent experiment called LWR-DEPUTY (coordinator, SCK.CEN). In this program, the main tests on Th-MOX consisted of additional fuels studies (microscopy, radial distributions of elements and isotopes) and radiochemical analyses. The objective of these analyses was to obtain a reliable experimental database for burn-up analysis and to evaluate changes in the heavy nuclide content:

• To optimize the dissolution and analysis strategies

• To establish the first dataset on heavy nuclide and fission product content in irradiated Th-MOX to assess the overall uncertainties

• To use this dataset in a benchmark analysis program

The OMICO Project [4] was conducted from 2001 to 2007. Its scope included the study and modeling of the influence of microstructure and matrix composition on Th-MOX fuel in-pile behavior in normal PWR conditions. The following tasks were undertaken:

• Fabrication of the Th-MOX fuels at the JRC-ITU

• Irradiation in the “CALLISTO” PWR loop in BR2, representing real PWR conditions; the burn-up achieved at the end of this project was about 13 GWd/ tHM

• Nondestructive examinations (gamma-spectrometry, visual examinations) and

microstructure studies

It should be noted that the pins were instrumented for pressure and fuel temperature determination. The test matrix was such that the Th-MOX could be compared with U-MOX and UO2 fuels. Another test parameter consisted of the fabrication process (homogeneous versus heterogeneous powder mixtures). The results of the temperature/pressure readings were primarily used to benchmark computer code models for Th-MOX fuels behavior in the first stage of their life.

Besides the irradiation, fuel characterization was performed, including thermal diffusivity measurements, and the results were published [4, 9]. The results show a similar thermal conductivity for (nonirradiated) Th-MOX as compared to U-MOX. In the LWR-DEPUTY [5] project, selected samples of the OMICO and THORIUM CYCLE programs were extensively studied to provide experimental datasets suitable for evaluating their in-pile performance. The experimental data were the basis of a benchmark exercise on the Th-MOX fuel pin irradiated at the NPP KWO to investigate the qualification of the numerical tools and software packages. A scoping study of the leaching behavior was also conducted. In addition to the experimental work, steady-state and transient analyses were performed for different PWR designs fueled completely or partially with Th-MOX fuel. An assessment of steady-state parameters (reactivity, shutdown margin, and reactivity feedback coefficients) has been performed in comparison with UO2. All feedback coefficients are favorable for a safe operation under steady-state conditions. A comparative analysis of control rod ejection scenarios has also been performed, and it was found that the maximum values obtained for fuel and clad temperature and maximum fuel enthalpy are in line with the acceptance criteria for the current generation PWRs.

After 10 years of research sponsored through the EURATOM programs, the following conclusions can be drawn regarding the behavior of Th-MOX fuel in LWR conditions:

• Th-MOX has great potential and its fabrication as an oxide fuel is feasible

• Even at a laboratory-scale production route, Th-MOX shows a good in-pile performance

• Know-how on Th-MOX has increased, but

• Fuel performance obviously needs to be further improved before code calculations can predict specific Th-MOX behavior

As a general conclusion, the results of these experiments have shown that Th-MOX behaves in a comparable way (even better in some aspects) to MOX, and that licensing Th-MOX in a LWR should not be problematic, although more experimental data on fuels representative of the future commercial fuels would be needed. Experimental data also demonstrate that Th fuels will be more resistant to corrosion than U fuels in the case of spent fuel geological disposal.

The Molten Salt Reactor

The MSR, which incorporates the reprocessing on line and needs no specific Th fabrication, adds the benefits of Th without its main challenges. In particular, breeding may be achieved over a wide range of neutron energies, which is not the case for the U-Pu cycle.

Under the European Framework Programs, conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts open promising possibilities to exploit the 232Th-233U cycle. In addition, they can also contribute to significantly diminishing the radiotoxic inventory from present reactor spent fuels, in particular by lowering the masses of transuranic elements. Finally, if required because of expansion of nuclear electricity generation breeding beyond the iso-generation could be achieved. With the Th-U cycle, doubling times values are only slightly higher than those predicted for solid-fuel fast reactors working in the U/Pu cycle (in the range 40–60 years). The characteristics of different launching modes of the MSFR with a thorium fuel cycle have been studied, in terms of the safety, proliferation, breeding, and deployment capacities of these reactor configurations [10].

Between Framework Programmes 5 and 7, several projects (“MOST”, “ALISIA”, “EVOL”) were conducted, and promising developments and results were obtained in particular in the following areas:

• Conceptual design studies

• Safety developments, in particular, to study the residual heat extraction; tests with liquid salts have been undertaken to prove the ability of the cold plug system to act as a security valve on the loop circuit

• Fabrication of the salt mixture (LiF-NaF-KF) to be used in the French molten salt loop (FFFER project) has been achieved

• Experimental investigation of physicochemical properties of fluoride salts

• Experimental tests of the metallic-phase extraction process;

• Corrosion studies and experiments (this remains one of the main challenges for the development of the reactor system)

Finally, it should be noted that the MSR with its Th cycle is one of the six reference systems selected for R&D collaboration in the framework of the Generation IV International forum. The main contributors are the European partners, supported by Russia as observer.


Since the early 1970s, studies and experimental projects have been undertaken in Europe to examine the potential of Th-based fuels in a variety of reactor types. These projects have all been successful from a scientific point of view, but not all were followed up relative to the overall development of nuclear industry in Europe. High-temperature reactors (HTRs), although very well suited for Th use, have not been deployed to the benefit of LWRs. Results on the use of Th matrices in Th-MOX fuels in LWRs are encouraging, but still need demonstration at a larger scale in commercial conditions. Finally, the probably most efficient use of Th would be in a salt, to feed MSRs. Conceptual studies and related experimental programs are under way.

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