M. Ugajin

Chemical form of the solid fission products in (Th, U) O2 simulating high burnup

The chemical form of the solid fission products has been studied for (Th0.81U0.19) O2 simulating 21.5% FIMA in an HTGR environment. Experiments have been performed with X-ray diffraction, electron-probe microanalysis, ceramography and hardness measurement. The results showed that fission-product phases of two types, Mo-Ru-Pd and (Ba, Sr)(Zr, Ce) O3, are present in the simulated fuel pellet. The fuel matrix comprises (Th1-xUx, Zr, Ce, RE) O2 with an x value of 0.067(RE = Nd, La, Pr, Y, Sm). Dissolution of the rare earths (RE) and the residual Zr plus Ce in (Th0.933U0.067) O2 was accompanied by contraction of the unit cell of the oxide matrix. Reaction behavior in the selected fission product system BaO-ZrO2-Nd2O3 was also investigated. The results showed that in the presence of BaO, Nd2Zr2O7 is converted to barium zirconate: at 1630°C, Nd2Zr2O7 + 2 BaO → 2 BaZrO3 + Nd2O3. This fact, combined with thermochemical assessment, confirms the relative stability of (Ba, Sr)(Zr, Ce) O3 against Nd2Zr2O7 in the simulated (Th0.81U0.19) O2. From these results and fission product inventories, it is inferred that the chemical state of high-burnup ThO2 is very similar to that of (Th0.81U0.19) O2.

Oxygen potentials of (th, u)o2+x solid solutions

The oxygen potential and the nonstoichiometry (x) for (Th1−yUy)O2+x were measured in situ at 1000° 1200°C using a solid electrolyte oxygen sensor and a thermobalance, respectively. The results showed that the oxygen potentials of (Th1−yUy)O2+x are not a function only of the U valence in the investigated range of 0.05⩽y⩽0.20. The oxygen potentials at 1200°C decrease negatively with increasing thorium content at a given U valence. The calculated activity coefficients of urania in the solutions indicate an increasing positive deviation from ideality with increasing U valence.

Irradiation behavior of high uranium-density alloys in the plate fuels

Abstract The U 6 Mn and U 6 Fe 0.6 Mn 0.4 alloys were irradiated to ∼54% of the initially contained 19.6% 235 U at around 190°C for 216 days. The volume swelling of the miniplates with such fuel dispersions in an Al matrix was reduced to the previously reported U 6 Fe plate. Irradiation tests using U 6 No 0.6 Fe 0.4 and U 3 Si 0.8 Ge 0.2 proved that the cladding restraint is more effective to suppress the gas-bubble growth in the foil-type than in the dispersion-type plates. The fuel-aluminium reaction was also investigated.

Thermodynamic properties of Th0.80U0,20O2 + x solid solution

Oxygen-potential (ΔGO2) measurements employing a thermogravimetric method have been performed for Th0.80U0,20O2 + x. A complete set of data is presented at 1273–1473 K in the ranges 2.000 ≲ OM ≲ 2.024 and −95 ≲ ΔGO2

Stability of the simulated fission-product phases in (Th, U)O2

−32 kcal/mol. Partial molar entropies and enthalpies of solution of oxygen in the mixed oxide were derived from the temperature variation of ΔGO2. Vapor pressures over Th0.80U0,20O2 + x at 2000 at 2300 K were calculated from our experimental ΔGO2 data and the known free energies of formation for gaseous and condensed oxides. It is predicted that with an increase in O/M ratio the vapor pressure of UO3(g) increases rapidly while maintaining an extremely lower pressure of ThO2(g).

Contribution to the study of the system UPuz.sbnd;Wz.sbnd;C

The oxygen-potential dependence of the stability of fission-product phases has been studied for (Th0.81U0.19)O2 simulating 21.5% FIMA. In reducing environments Mo-Ru-Pd alloy and (Ba, Sr)(Zr, Ce)O3 exist stably in the fuel matrix, whereas in oxidizing environments (Ba, Sr)MoO4 and Nd2(Zr, Ce)2O7 become more stable than the perovskite-type zirconate, with oxidative loss of Mo from the alloy. The oxygen-potential threshold for such a change in the chemical state of the fuel is in the range ΔG(O2) = −62.0 to −69.6 kcal/mol at 1500°C. The threshold ΔG(O2)-value will coincide with that for the onset of Mo oxidation in the alloy; i.e., − −65 kcal/mol at 1500°C. The possible role of molybdenum in an oxide fuel pin is briefly considered on the basis of oxidation behavior of the molybdenum and the fuel matrix.

The phase reaction in the UC-W system

M-W-C alloys (M = U and/or Pu) have been investigated with the methods of X-ray diffraction and metallography. A new compound PuWC2 was found; its structure is UWC2-type orthorhombic with lattice parameters a = 5.621 ± 0.003 A, b = 3.245 ± 0.002 A, c = 10.877 ± 0.007 A. Phase relations in the Pu-W-C system were also studied and the constitutional diagram at 1400 °C is proposed. The occurrence of the orthorhombic PuWC2-UWC2 solid solutions, (U, Pu)WC2, has been confirmed in the entire range of composition. Alloys of compositions UWC1.64−2.20 were examined to determine the phase boundaries at 1700 °C. The results show that the orthorhombic phase has a narrow stoichiometry range with approximate limits of 0.00 ⩽ x ⩽ 0.10 in UWC2−x; the cell size slightly decreases with carbon dissolution in the defect lattice. The existence of the UMoC1.70-type monoclinic UWC1.75, previously denoted η-UWC2, has been established, with lattice parameters a = 5.6248 ± 0.0012A, b = 3.2433± 0.0008 A, c = 11.650 ± 0.002 A, β = 109.60 ± 0.01°. X-ray diffraction powder data are presented for PuWC2, UWC2 and UWC1.75.

Behavior of neutron-irradiated U3Si

Abstract The phase reaction in the UC-W system has been investigated by metallography, X-ray diffraction, microhardness measurement and electron-microprobe analysis. The results show that in the system a peritectic four-phase reaction takes place: UC + W / ag η - UWC 2 + liquid ( at 2150 ± 20 ° C ), where η-UWC 2 denotes a low carbon form of UWC 2 . The compositions of the peritectic liquid and of the peritectic point are determined to be near 48 U/34 C/ 18 W (at%) and 40 U/40 C/20 W (at%), respectively. The phase diagrams are presented for the U-C-W ternary and UC-W quasibinary systems.

Reactions between UZr alloys and nitrogen

Abstract The behavior of U 3 Si-based alloys has been studied under neutron irradiation. Maximum burnup reached ∼62% of the initially contained 19.6% 235 U and irradiation temperatures were in the approximate range of 190–280°C. Postirradiation examinations revealed the following. The dimensional instability of the high uranium-density fuels is attributed to the radiation-induced amorphization and plastic deformation. The occurrence of amorphization is suggested by the liquid-like behavior of U 3 Si under irradiation at temperatures well below the melting point of crystalline material. The accelerated swelling of U 3 Si due to the large fission-gas-bubble growth can be suppressed by the cladding restraint. The reaction layer is formed at the U 3 Si Al interface. The thickness of the reaction layer of surface-oxidized U 3 Si is significantly reduced in comparison with that of non-oxidized U 3 Si. This reduction in thickness is caused by the thin film of UO 2 that has been formed at the surface of the U 3 Si by oxidation.

Contribution to the study of the UCsMoIO system

Abstract The reactions of UZr alloys with nitrogen were studied at temperatures ranging from 873 to 1273 K. Electron probe microanalysis, wide angle and microbeam X-ray diffraction techniques were applied for the identification of reaction products. Adherent multi-layered scales were formed during the anneals under nitrogen pressures of 0.19 and 20 kPa. They were mainly composed of U 2 N 3 , ZrN and α-Zr(N). The scale thickness and morphology depended on the alloy compositions and temperatures.