Fuel cycle strategies and plutonium management in Europe

An overview of current nuclear power generation and fuel cycle strategies in Europe is presented, with an emphasis on options for the management of separated plutonium in the medium to long term. Countries which have opted for reprocessing of spent fuel have had to contend with increasing inventories of separated plutonium. Of the various potential options for utilisation or disposition of these stockpiles, only light water reactor (LWR) mixed-oxide (MOX) fuel programmes are sufficiently technologically mature to be fully operational in several European countries at present. Such reprocessing-recycling programmes allow for a stabilisation of the overall separated plutonium stocks, but not for a significant reduction in the stockpile. Moreover, the quality of recycled plutonium decreases at each potential step of re-irradiation. Therefore, optimised or new ways of managing the plutonium stocks in the medium to long term are required. In the present overview we consider the most promising options for reactor utilisation of plutonium in both near-term future reactor and Generation IV systems.     

Synthesis,structure, and characterization of the thorium zirconolite CaZr1-xThxTi2O7 system

A series of zirconolite ceramics with composition CaZr_1-xTh_xTi₂O₇ (Δx = 0.10) were reactively sintered at 1350°C for 20 h, in air (0 ≤ x ≤ 0.60) and 5% H₂/N₂ (0 ≤ x ≤ 0.40). A sample with composition corresponding to x = 0.20 was also produced by hot isostatic pressing (HIP) at 1300°C and 100 MPa for 4 hours. Th⁴⁺ immobilization was most readily achieved under oxidizing conditions, with Th⁴⁺ preferentially incorporated within a pyrochlore-structured phase in the range 0.10 ≤ x ≤ 0.50, yet formation of the zirconolite-4M polytype was not observed. We report the novel synthesis of single-phase pyrochlore with nominal composition CaZr_0.40Th_0.60Ti₂O₇ when targeting x = 0.60. Th⁴⁺ incorporation under reducing conditions produced a secondary Th-bearing perovskite, comprising 24.2 ± 0.6 wt% of the phase assemblage when targeting x = 0.40, alongside 8.8 ± 0.3 wt% undigested ThO₂. Under reducing conditions, powder XRD data were consistent with zirconolite adopting the 3T polytype structure. The sample produced by HIP presented a nonequilibrium phase assemblage, yielding a major phase of zirconolite-2M alongside accessory Th⁴⁺-bearing phases ThTi₂O₆, ThO₂, and perovskite. These data highlight the efficacy of Th⁴⁺ as a Pu⁴⁺ surrogate, with implications for the formation of Zr-stabilized Th-pyrochlore phases as matrices for waste with elevated Th⁴⁺ content.     

Long Term Change in Level-Volume Relation of an Accountability Tank for Plutonium Nitrate Solution

A slight change in the level-volume relation for an accountability tank for a large amount of plutonium nitrate solution (PuN) was observed at the Plutonium Conversion Development Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC), Tokai Works. From the results of annual tank re-calibrations for the plutonium receiving tank from 1985 to 1992 using the incremental feed of nitric acid as the density standard, it became clear that the relation between the level and the volume changed slightly, and the rate of the change was a linear function of operating time. Also it became clear that the change was linear in relation to the level. In the PCDF, the cumulative change in the volume at the nominal level was evaluated to be 0.1% during 8 years' operation. It was also evaluated that the repeatability of the re-calibration is much better than 0.1%. A reasonable frequency of tank re-calibration is once every 5 years.     

Removal of Liquid and Solid Impurities from Uranyl Nitrate Hexahydrate Crystalline Particles in Crystal Purification Process

The purification behavior of uranyl nitrate hexahydrate (UNH) was investigated to evaluate the decontamination performance of liquid and solid impurities using a dissolver solution of mixed oxide (MOX) fuel in batch experiments. The UNH crystal recovered from the MOX fuel dissolver solution containing simulated fission products (FPs) was purified by a sweating and melt filtration process. Although the decontamination factors (DFs) of Pu, Cs, and Ba did not change in the sweating process, that of Eu increased with increases in temperature and time. These results indicate that liquid impurities such as Eu were effectively removed by the sweating method, but solid impurities such as Pu, Cs, and Ba were minimally affected in the batch experiments. On the other hand, the DF of Ba increased with 0.45 and 5.0 μ filters in the melt filtration process. Since Pu and Cs formed as Cs₂Pu(NO₃)₆ in the course of U crystallization and was accompanied with the UNH crystal, these behaviors were similar to each other. Although the DFs of Pu and Cs did not change with the 5.0 μ filter, it increased approximately twofold with the 0.45 μ filter. The particle size of Cs₂Pu(NO₃)₆ is relatively small and might pass through the 5.0 μ filter in the melt filtration process. The liquid impurities as Eu remained in the molten UNH crystal with some filters.     

Management of Plutonium Content based on Reactivity of Each Plutonium Isotope

It is strongly demanded to control reactivity for safety and steady operation of nuclear reactor. Fissile enrichment should be, therefore, accurately managed in order to carry out it. A concept of physical accounting method for adjusting fissile enrichment is very useful to manage plutonium content and to control reactivity of nuclear reactor where plutonium is used as main fissile material. The basis of the concept is shown in this paper together with the result of the application to typical prototype fast breeder reactor (FBR). The rationality for adopting the method is also quantitatively shown in this paper.     

Accurate Volume Measurement System for Plutonium Nitrate Solution

Abstract

An accurate volume measurement system for a large amount of plutonium nitrate solution stored in a reprocessing or a conversion plant has been developed at the Plutonium Conversion Development Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC) Tokai Works. A pair of differential digital quartz pressure transducers is utilized in the volume measurement system. To obtain high accuracy, it is important that the non-linearity of the transducer is minimized within the measurement range, the zero point is stabilized, and the damping property of the pneumatic line is designed to minimize pressure oscillation. The accuracy of the pressure measurement can always be within 2 Pa with re-calibration once a year. In the PCDF, the overall uncertainty of the volume measurement has been evaluated to be within 0.2%. This system has been successfully applied to the Japanese government's and IAEA's routine inspection since 1984.     

Study on FBR core concepts to increase proliferation resistance of plutonium in LWR–FBR transition period

The possibility to enhance proliferation resistance of discharged plutonium in fast breeder reactor (FBR) has been investigated in terms of reactor core-design aspects. The provisional target for proliferation resistance measures based on Saito's attractiveness (ATTR) is defined. It is found that a few percent of plutonium loading and/or Am/Cm loading, which come from various types of spent fuel, might satisfy the provisional target with a minimum impact on the core neutronic performances. On Am/Cm loading core, decay heat constraints for fuel-handling aspects are found to be important and should be considered in design. There is not significant change on the current developing scenarios for light water reactor–FBR transition period by applying the measures based on Saito's ATTR. It is found that applying Kimura's proposal, 15% ²³⁸Pu content requires about 7% MA loading fraction both in the core and the blanket, and it only can be applied at limited case. The period to consume minor actinides is shorter than to consume plutonium.     

High temperature X-ray diffraction study of the kinetics of phase separation in hypostoichiometric uranium–plutonium mixed oxides

Uranium–plutonium mixed oxides incorporating high amounts of plutonium are considered for future nuclear reactors. For plutonium content above 20%, a phase separation occurs, depending on the temperature and on the oxygen stoichiometry. Here, using an in situ fast X-ray diffraction device dedicated to radioactive materials, we evidenced a phase separation occurring during rapid cooling from 1773 K to room temperature at the rates of 0.05 and 2 K s⁻¹ for two (U_1−yPu_y)O_2−x compounds, with y = 0.28 and 0.45, under a reducing atmosphere. Optical microscopy reveals that the cooling rate impacts the microstructure of the fuel pellet by inducing severe macroscopic cracks. These results are important for using uranium–plutonium mixed oxides with high plutonium content as nuclear fuels. Considering the associated issues, they dictate a cautious attitude when defining their conditions of fabrication.