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.     

(U,Gd)O2芯块试验与工业生产

文章介绍在宜宾核燃料元件厂(YFP)生产线上进行(U,Gd)O2芯块工业规模的生产试验及产品合格性鉴定,对试验结果进行讨论和评价.结果表明:YFP的(U,Gd)O2芯块生产线完全具备工业生产能力,并实现(U,Gd)O2芯块制造的国产化.     

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.     

Electrical conductivity of near-stoichiometric (U,Gd)O2 solid solutions

The electrical conductivity of near-stoichiometric (U, Gd)O₂ solid solutions containing gadolinia up to 14 mol% was measured by four probes dc technique. The addition of gadolinia to urania enhanced the p-type electrical conductivity in the temperature range from room temperature to 1000 K. The electrical conductivity was able to be interpreted in terms of hopping of localized holes between U⁵⁺ and U⁴⁺ ions in the solid solutions. The experimental results presented in this study were consistent with the behavior predicted by the adiabatic small polaron theory.     

Electrochemical Reduction of (U,Pu)O2 in Molten LiCl and CaCl2 Electrolytes

The electrochemical reduction of UO₂-PuO₂ mixed oxides (MOX) was performed in molten LiCl at 923 K and CaCl₂ at 1,123 K to evaluate the behavior of the plutonium quantitatively and to define the optimum conditions for the electrochemical reduction of those materials.

In LiCl, excess deposition of lithium metal can be avoided and the MOX was smoothly reduced at ?0.65 V vs. Bi-35 mol% Li reference electrode. The reduction ratio calculated from the mass change of the samples taken during the electrochemical reduction and the ratio evaluated by gas-burette method were in good agreement. The cathodic current efficiency remained 30–50% mainly due to the deoxidation of tantalum cathode basket. Although dissolution of plutonium and americium into the electrolyte was found by the chemical analysis, the dissolved amount was negligible and had no immediate influence on the feasibility of the electrochemical reduction process.

In CaCl₂, reduction of the MOX occurred in whole range of the tested cathode potential (?0.15 V to ?0.40 V vs. Ca-Pb reference electrode). The cathodic current efficiency was around 30%. Although the MOX was completely reduced at ?0.25 V, the reduction was interrupted by formation of the surface barrier made of the reduced material and the vacancy between the reduced and the non-reduced areas at ?0:30 V. Plutonium and americium dissolved also into the CaCl₂ electrolyte to slightly higher concentrations than those observed in LiCl electrolyte. The analyses for the reduction products showed that the amount of those actinides lost from the cathode was much larger than that found in the electrolyte, probably due to the formation of mixed oxide precipitate.     

Fabrication and characterization of (U,Am)O2−x transmutation targets

A novel dust-free route for the preparation of (U, Am)O_2?x targets has been demonstrated using a combined sol–gel and single- or double Am-infiltration process at the Minor Actinide Laboratory (MALAB) of the Institute for Transuranium Elements (ITU). Samples with 10 and 20 mol% of Am were prepared. For both Am concentrations, a single cubic phase material with a fluorite structure was observed by X-ray diffraction. X-ray absorption spectroscopy was carried out to characterize the chemical state of the metal atoms and their local crystallographic environment. The U(IV) and Am(III) valence states are predominant and the O/Am ratio is ～1.6 for both Am contents. For the 20 mol% Am, EXAFS reveals an expansion of the Am–O (2.43 Å) bond length beyond the metal–oxygen bond length in both AmO₂ (2.31 Å) and UO₂ (2.35 Å).     

Chlorination of UO2 and (U,Zr)O2 solid solution using MoCl5

Conversion of UO₂ and (U_0.5Zr_0.5)O₂ solid solution into chlorides by MoCl₅ was performed in order to confirm the applicability of the chlorination by MoCl₅ to a pretreatment of fuel debris by pyrochemical methods. Chlorination of (U_0.5Zr_0.5)O₂ powders and dense pieces was successfully achieved at 573 and 773 K, respectively, based on the following chemical reaction: 2(U_0.5Zr_0.5)O₂ + 4MoCl₅ = UCl₄ + ZrCl₄ + 4MoOCl₃. Rough separation of MoCl₅, ZrCl₄ and MoOCl₃ from UCl₄ was achieved by volatilization under temperature gradient. From these results, fundamental feasibility of the chlorination method using MoCl₅ as a pretreatment of fuel debris was shown.     

In-Pile and Out-of-Pile Grain Growth Behavior of Sintered UO2 and (U,Gd)O2 Pellets

Grain growth behavior of UO₂ and (U, Gd)O₂ fuel pellets was investigated with the data from the out-of-pile isothermal heating experiments and the irradiation test at the Halden Boiling Water Reactor. The laboratory data gave best-fitted equations by employing the following fourth power rate equations :

UO₂ : D²-D⁴ ₀=3.79×10¹⁸ exp(-142,000/RT)t,

(U, Gd) : D²-D⁴ ₀=4.98×10¹⁷ exp(-140,000/RT)t,

where, D ₀ and D are initial and final three-dimensional diameters (μm), respectively, R the gas constant (=1.987 cal/mole/K), T the absolute temperature (K) and t the time (h) (gadolinia content: 3~10%, temperature range: 1,700~2,000°C).

The calculated grain diameter with the above equations revealed an overestimation on specimens which involved noticeable fission gas bubbles on their grain boundaries. It was demonstrated that the in-pile grain growth model, as was given in the following equation, which took account of the retarding effects of growth by precipitated intergranular bubbles could describe the grain growth of the irradiated samples :

where f: Grain boundary fractional coverage (-).     

Sol-gel extrusion process for fabrication of (Th,U)O2 recycle fuel

An extrusion process based on sol-gel derived paste has been developed for the production of thoria recycle fuel as an alternative to the conventional powder-compaction/sintered-pellet route. Crack-free, high-density (9.7 Mg/m³) extruded slugs have been fabricated from sol-gel pastes prepared from ThO₂ powder (denitrated at 600°C) having a moisture content of ~ 16%, and mixed with phenolic resin. The extruded slugs were finally sintered at 1600°C. The effects of thorium-nitrate denitration temperature and organic binder addition on the quality of the fuel slugs are discussed in the paper.     

Fabrication of high-density UO2 and (U0.75Pu0.25)O2 by hot pressing

The hot pressing behavior of hyperstoichiometric UO₂ and (U, Pu)O₂ powders has been evaluated. Specimens with densities in excess of 99% $\text{ρ}{}_{\mathrm{th}}$ can be fabricated with this technique at temperatures of 1000°C or less.     

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U_1?xZr_xO₂ and (U,Pu)_1?xZr_xO₂ were carried out in 6 M HNO₃ at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10^?5 down to 10^?10 mol cm^?2 min^?1 as x increased from 0 to 0.95. From these findings, dissolution with HNO₃ is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO₃ at 353 K is assumed. Application of complexing acids, such as mixture of HNO₃ and HF, should be considered to increase the dissolution rate of the Zr-rich part.     

Behavior and Removal of Impurity Sulphate in Sol-Gel Process for ThO2 and (Th,U)O2 Microsphere Fabrication

The behavior of impurity sulphate, fed from source Th(NO₃)₄, in a sol-gel process for fabricating (Th, U)O₂ microspheres is studied. In sols, almost all the sulphates are adsorbed on colloid particles regardless of U content U/M (M:Th+U). However, the sulphate in gelmicrospheres is easily washed out for U/M higher than 10mol/₀, while not for lower U/M; especially, the sulphate in ThO₂ gel-microspheres is difficult to be removed even by long-time washing. The sulphur remaining in the gel-microspheres is able to be removed out by heating in air or steam at 1,000°C for 3h; as for ThO₂, slight residue is found but not after 1,300°C heating. On the other hand, heating them in Ar-4%H₂ atmosphere, even at 1,500°C, allows most of the sulphur to remain as an unidentified M-O-S compound; during the heating, the sulphur is gradually released by the reaction with H₂ to form H₂S.

The study finds optimum conditions for removal of the impurity sulphate in the sol-gel process.     

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th,²³³U)O₂ mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with ²³²U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO₂ is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O₂ pellets. In this study, fabrication of (Th,U)O₂ mixed oxide pellets containing 3–5 wt.% UO₂ was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.     

萃取色层分离/ICP-AES法测定钆铀氧化物中的杂质

研究了钆铀氧化物中29种杂质元素的测定。采用CL-TBP从硝酸溶液中萃取色层分离基体铀，所得含杂质及钆的淋出液以电感耦合等离子体原子发射光谱仪采用多组分图谱拟合法(MSF)测定。研究了MSF模型的建立与考查、不同杂质浓度下的回收率等。结果表明，方法对钆铀样品中29种杂质元素，即Al、Ba、Ca、Co、Cr、Cu、Fe、Li、Mg、Mn、Mo、Na、Ni、Ti、Zn、Ta、W、Th、Bi、Pb、Sn、V、In、Ag、B、Cd、Sm、Eu、Dy的测定准确可靠。对于其它杂质含量相当的铀／钆化合物，通过适当的处理和转化，也可采用本方法测定。