Densification behaviour and sintering kinetics of ThO2-4%UO2 pellet

ThO₂-?4% ²³³UO₂ fuel will be the driver fuel for the forthcoming Advanced Heavy Water Reactor (AHWR) in India. Densification behaviour such as shrinkage and shrinkage rates of the green pellets of ThO₂-4wt.% UO₂ (natural ‘U’) fabricated by Coated Agglomerate Pelletization (CAP) process were studied using a vertical dilatometer at different heating rates. Activation energy of sintering, ‘Q’, was estimated in the initial stages of sintering by continuous rate of heating (CRH) technique as proposed by ‘Wang and Rishi Raj’ and ‘Young and Cutler’. The sintering mechanism was identified to be as the grain boundary diffusion (GBD) and the average ‘Q’ value obtained by these two methods were found to be 350 ± 16 kJ/mole and 358 ± 5 kJ/mole, respectively.     

Production and investigation of thorium-containing fuel compositions

A technology has been developed for obtaining fuel tablets with the compositions (U, Th)O₂, (U, Th, Ca)O₂, and (U, Th)O₂+MgO by combined precipitation of uranium, thorium, magnesium, or calcium components from inert solutions, followed by heat treatment of the powders, compression into pellets, and sintering of the pellets. Work on optimizing the technological processes for obtaining fuel pellets so as to obtain good pellet quality was performed. The effect of the properties of the precipitates and powders, fabricated using different technological regimes on the properties of the finished objects was studied. The work includes detailed investigations of powders (x-ray phase analysis, electron-microscopic investigation) and sintered fuel tablets (change in the geometric dimensions as a result of sintering, determination of the density, and study of the microstructure). The behavior of fuel compositions (U, Th)O₂ and (U, Th)O₂+MgO in contact, with coolants under conditions where the fuel elements become unsealed was studied: with water at 300°C and sodium at 700°C. 3 figures, 3 tables, 6 references. State Science Center of the Russian Federation-A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 5, pp. 346–353, May, 2000.     

Investigation of ThO2 and (U, Th)O2

The effect of the properties of ThO₂ and (U, Th)O₂ powders, prepared with different technological regimes, on the properties of the finished items is investigated. The work includes detailed investigations of ThO₂ and (U, Th)O₂ powders (x-ray phase analysis, electron-microscope investigation) and sintered fuel pellets (determination of density, study of microstructure, thermophysical investigations). The temperature dependences of the crystal lattice parameters and the sizes of the crystallites in ThO₂ and (U, Th)O₂ powders with different UO₂:ThO₂ ratio are obtained. The temperature dependences of the thermal conductivity of sintered ThO₂ and (U, Th)O₂ pellets with different UO₂:ThO₂ ratio are studied.     

Characterization of (Th,U)O2 pellets made by advanced CAP process

Coated Agglomerate Pelletization (CAP) process is being developed by Bhabha Atomic Research Centre (BARC) for the fabrication of ThO₂-UO₂ mixed oxide fuel pellets. In order to improve the microstructures with better microhomogeneity, a study was made to modify the CAP process. The advanced CAP (A-CAP) process is similar to the CAP process except that the co-precipitated powder of mixed oxide, ThO₂-30%UO₂ or ThO₂-50%UO₂, is used for coating instead of U₃O₈ powder. The choice of ThO₂-UO₂ powders as the coating material is advantageous compared to U₃O₈, since the presence of large quantities of ThO₂ in UO₂ powder gives better self-shielding effect. In this paper, ThO₂ containing 4%UO₂ (% in weight) was prepared by the A-CAP process. Property measurements including microstructure and microhomogeneity were made by optical microscopy, scanning electron microscopy (SEM), electron probe microanalysis (EPMA), etc. It was found that the pellets sintered in air at 1400 °C showed a duplex grain structure and those sintered in Ar-8%H₂ at 1650 °C showed a very uniform grain structure with excellent microhomogeneity.     

A non destructive technique for determination of U and Pu content in MOX fuel pins for fast reactors

MOX fuel pins containing both U²³³O₂ and PuO₂ have been fabricated for making an experimental subassembly for irradiation in Fast Breeder Test reactor (FBTR) at Kalpakkam, India. This unique composition of the fuel pin is chosen to simulate the thermo-mechanical conditions of the upcoming Prototype Fast Breeder Reactor (PFBR) in the existing Fast Breeder Test Reactor. Since the fertile matrix is natural UO_2, it was difficult to monitor the percentage of U²³³O₂ through chemical methods and neutron assay methods. During the fabrication of MOX fuel pins at Advanced Fuel Fabrication Facility; Bhabha Atomic Research Centre, Tarapur, Passive Gamma Scanning (PGS) was employed as one of the characterisation tools for verifying the fuel composition. PGS was found to be effective in estimating the percentage composition of both U²³³O₂ and PuO₂ and also in ensuring the uniform distribution of the fissile material in MOX fuel pins. PGS is also found effective in monitoring the correct loading of natural UO₂ insulation pellets and MOX fuel pellets in welded MOX pins.     

Mechanical properties and XRD studies of silicon carbide inert matrix fuel fabricated by a low temperature polymer precursor route

The mechanical properties of silicon carbide (SiC) inert matrix fuel (IMF) pellets fabricated by a low temperature (1050 °C) polymer precursor route were evaluated at room temperature. The Vickers hardness was mainly related to the chemical bonding strength between the amorphous SiC phase and the β-SiC particles. The biaxial fracture strength with pre-notch and fracture toughness were found to be mostly controlled by the pellet density. The maximum Vickers hardness, biaxial fracture strength with pre-notch and fracture toughness achieved were 5.6 GPa, 201 MPa and 2.9 MPa m^1/2 respectively. These values appear to be superior to the reference MOX or UO₂ fuels. Excellent thermal shock resistance for the fabricated SiC IMF was proven and the values were compared to conventional UO₂ pellets. XRD studies showed that ceria (PuO₂ surrogate) chemically reacted with the polymer precursor during sintering, forming cerium oxysilicate. Whether PuO₂ will chemically react in a similar manner remains unclear.     

Neutron Behavior in Cluster-Type Fuel Lattices, (I)

Lattice parameters φ²⁸, φρ²⁵, ρ²⁸ and C^* were measured on cluster-type fuel lattices of the ATR (Advanced Thermal Reactor) by using a two region critical facility (D₂0-cluster test and H₂O-rod driver regions). Their dependence on lattice pitch, coolant-void ratio and fuel composition (whether UO₂ or PuO₂-UO₂) have been made clear by this experiment.

A foil handling technique has been developed for determining the lattice parameters of the Pu0₂- UO₂ fuel pins, and the resulting measurement errors are almost as small as those obtained on the U0₂ fuel pins.

The effects of the Cd-filter and of the presence of UO₂ buttons in the measurement of ρ²⁸ and χ²⁵ were studied experimentally and correction factors have been determined.

A method of observing the spatial distribution of the γ-ray source in an activated foil has been developed, and the relation between the spatial distribution and the coincidence counting efficiency of the foil has been examined.     

Leaching behaviour of unirradiated high temperature reactor (HTR) UO2-ThO2 mixed oxides fuel particles

The dissolution of different mixed oxide (U, Th)O₂ particles under reducing conditions has been studied using a continuous flow-through reactor. The U/Th ratio seems to have no or little influence on the normalised leaching rate of thorium or uranium, The release rate of uranium from the outer surface of a Th rich matrix seems to follow the behaviour of pure UO₂ even though U is a minor component in these phases and the dissolution rate of Th is much lower. After long time U concentrations will become depleted at the solids surface and it will be expected that U release rates will become controlled by the release rates of thorium (rates at neutral pH < 10⁻⁶ g m⁻² d⁻¹). Under reducing conditions, the matrix of HTR fuel particles presents significant intrinsic radionuclide confinement properties.     

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens’ theory and reported thermal conductivities of unirradiated (U, Pu) O₂ and irradiated UO₂ pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

Thermophysical properties of (U,Y)O2

We prepared polycrystalline pellets of (U,Y)O₂, containing YO_1.5 up to 11 mol.%. We performed indentation tests on the pellets, and evaluated the Young’s modulus and hardness. We measured the heat capacity and the thermal diffusivity, and evaluated the thermal conductivity. We succeeded in evaluating the effect of Y content on the thermophysical properties of (U,Y)O₂. We revealed that the Young’s modulus, hardness, and thermal conductivity of (U,Y)O₂ decreased with increasing the Y content.     

Fabrication of dense (Th,U)O2 pellets through microspheres impregnation technique

ThO₂ microspheres were prepared by internal gelation process using a pre-boiled hexamethylenetetramine (HMTA), urea solution. The microspheres were characterized with respect to tap density, specific surface area and pore size distribution. An indigenously designed and fabricated apparatus was used for the impregnation of uranium in thoria microspheres. The loading of uranium was found to vary with the concentration of uranyl nitrate solution, operational vacuum and the time of impregnation. These process conditions were optimized to obtain soft (Th,U)O₂ microspheres containing 3-4 mol% of uranium, which are readily amenable for pelletization. The green pellets could be sintered to ∼96% of T.D. by heating in air up to 1350 °C for a period of 2-4 h. The polished surface of the fractured pellets showed a smooth surface without any berry structure. The shrinkage behaviour of the pellets was also studied in air using a dilatometer. The SEM studies of the pellets indicated a uniform microstructure with average grain size of 1 μm. The elemental scanning by the EDX method showed the uniform distribution of uranium in the microspheres and pellets.     

Thermal and mechanical properties of (U,Er)O2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er₂O₃ to all UO₂ pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of ²³⁵U enrichment. When utilizing the (U,Er)O₂ fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U_1−xEr_x)O₂ (0 ? x ? 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O₂. All Er completely dissolved in UO₂, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O₂ decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O₂ fuels in LWRs.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.     

Dimensional Changes in Irradiated (Th,U)O2

Effects of irradiation on the dimension and microstructure in (Th,U)O₂ pellets were examined by measurements of lattice parameter and bulk density changes, and observations of pore structures. The concentrations of fission-induced defects and the damage volume were estimated by a simple model. Both macroscopic and microscopic dimensional changes were found to increase initially with fission dose and then fall off. The difference between macroscopic and microscopic ingrowths increased with dose, suggesting that fission-induced interstitials would cluster or go to sinks and the concentration of vacancies would be in excess of that of interstititials. The damage volume for vacancies was estimated to be about 1x10^?22m³·fiss.^?1, and almost agreed with that for fission Xe release. Observations of the pore structure indicated that the volume fraction of pores smaller than 2–3 μm decreases with irradiation and the distribution of pore size shifts toward the larger side.     

Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of ²³⁸U, the fertile isotope in standard UO₂ fuel, are replaced by ²³²Th, then a greater yield of new fissile material (²³³U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 10³ years. Two innovative fuel types named S90 and S20, ThO₂ mixed with 90% and 20% enriched UO₂ respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated. Finally, we discuss the economics of such strategies.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

Startup and burnup strategy for Th–U/U–Pu fuel cycles in an EM2 reactor

General Atomics (GA) is developing the Energy Multiplier Module (EM²) which is a compact gas-cooled fast reactor as one of candidates of the Generation-IV nuclear energy systems. In the EM² core, low enriched uranium is used as igniting fuel and depleted uranium is used for converting and burning. It indicates that EM² can maintain critical operation for more than 30 years without refueling. To further study the Th–U fuel cycle performance in the EM², two kinds of start-up strategies with Th–U (Th + ²³³U) and semi Th–U (Th + enriched ²³⁵U) are evaluated. Neutronics characteristics, such as the effective multiplicity factor (k_eff) and conversion ratio (CR) are analyzed from neutron usage point of view. The simulated results for the two kinds of fuels are compared with the U–Pu fuel from the design of GA. The analysis gives an insight into the pros and cons of U–Pu and Th–U fuel cycles in terms of the breeding capability and the discharged radio-toxicity. The breeding performance of the second generation EM² is also presented and compared with that of the first generation EM². It indicates that the multi-generation EM² can deepen the burnup and reduce the waste management pressure for each kind of fuel loading strategy.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.