A Thermodynamic study of fluorite/bixbyite equilibria in the UO-PuO-PuO system

Vinograd, V. L.*; Vauchy, R.

Journal of Nuclear Materials, 619, p.156244_1 - 156244_16, 2026/01

https://doi.org/10.1016/j.jnucmat.2025.156244

Thermodynamic properties of (UPu)O fluorite (FCC) and (UPu)O bixbyite (BCC) in the UO-PuO-PuO system are assessed by considering phase equilibrium constraints and data on the variation of oxygen to metal ratio (O/M) as a function of the chemical potential of O. Thermodynamically, both BCC and FCC are described as ordered solid solutions allowing for a decrease in the configurational entropy of their oxygen/vacancy distributions at the specific values of = -0.5 and = -0.375 (O/M = 1.5 and O/M = 1.625). With this approach, fluorite/bixbyite equilibria in PuO-PuO and in UO-PuO-PuO are reproduced well with little effort. Moreover, we show that a large manifold of experimental data on the UO-PuO-PuO system is consistent with the assumption that Pu/(Pu+U) ratios in individual phases remain equal to the total Pu/(Pu+U) ratio in the system, i.e., no inter-phase U/Pu-partitioning occurs.

Oxygen potential and oxygen diffusion data for guiding the manufacture of MOX fuel for fast neutron reactors

Vauchy, R.; Horii, Yuta; Hirooka, Shun; Akashi, Masatoshi; Sunaoshi, Takeo*; Nakamichi, Shinya; Saito, Kosuke

Journal of Nuclear Materials, 616, p.156115_1 - 156115_16, 2025/10

https://doi.org/10.1016/j.jnucmat.2025.156115

Effects of Al addition on Vickers hardness increase by thermal aging of Fe-Cr-Al alloys; Evaluation by systematic experiments, machine learning modeling, and first-principles calculations

Abe, Yosuke; Tsuru, Tomohito; Fujita, Yohei*; Otomo, Masahide*; Sasaki, Taisuke*; Yamashita, Shinichiro; Okubo, Nariaki; Ukai, Shigeharu

Journal of Nuclear Materials, 606, p.155606_1 - 155606_12, 2025/02

https://doi.org/10.1016/j.jnucmat.2025.155606

We investigated the effect of Al addition on the formation of phase in Fe-Cr-Al model alloys by thermal aging. The Vickers hardness tests and a machine learning model indicate that the formation of the phase is promoted by low Al additions and suppressed by high Al additions. First-principles calculations, which indicate that Cr-Al-vacancy pairs are more stable than Cr-Cr pairs and that including Al atoms during phase nucleation may be energetically advantageous. On the other hand, the formation of Al-Al pairs was very unstable. The formation of Al-Al pairs near the interface can be avoided when the amount of Al addition is small. However, it is inevitable when the amount of Al addition is significant, leading to the instability of the phase.

Systematic experimental and model-based evaluation of the synergistic effects of alloy composition and damage rate on the formation of Cr-rich precipitates in Fe-Cr-Al alloys under ion irradiation

Abe, Yosuke; Sasaki, Taisuke*; Yamashita, Shinichiro; Okubo, Nariaki; Ukai, Shigeharu

Journal of Nuclear Materials, 600, p.155271_1 - 155271_12, 2024/11

https://doi.org/10.1016/j.jnucmat.2024.155271

To investigate the formation behavior of Cr-rich precipitates (CrRP) in Fe-Cr-Al (ODS) alloys being developed as accident tolerant fuel cladding for light water reactors, 14 Fe-Cr-Al alloys with systematically varied Cr and Al compositions were irradiated with 10.5 MeV Fe at C at three damage levels. A three-dimensional atom probe analysis showed that the CrRP number density, volume fraction, and Cr concentration increase with increasing Cr composition, decreasing Al composition, and decreasing dose rate. The result of the multiple regression analysis on CrRP volume fractions indicates that in addition to the primary effects of these variables, there are several important interactions. It was also highlighted that to understand the dose rate effect on the CrRP formation behavior under neutron irradiation, it is useful to examine the irradiation time dependence, including the effective use of thermal aging data as a limit to the zero dose rate.

Oxygen potential measurement of UAmO at 1473, 1573, and 1673 K

Watanabe, Masashi; Yokoyama, Keisuke; Vauchy, R.; Kato, Masato; Sugata, Hiromasa*; Seki, Takayuki*; Hino, Tetsushi*

Journal of Nuclear Materials, 599, p.155232_1 - 155232_5, 2024/10

https://doi.org/10.1016/j.jnucmat.2024.155232

Oxygen potential data of UAmO were measured at 1473, 1573, and 1673 K by thermogravimetry. In UAnO, where An stands for Pu or Am, and for a given value of y and Oxygen/Metal ratio, the oxygen potential of UAmO is higher than that of UPuO. The valence of cations in the hypostoichiometric region is similar to that of Nd-doped UO. At the stoichiometric composition, it is estimated to be Am, U, and U (for charge compensation of Am). The experimental data were analyzed using a defect chemistry model, and a relationship connecting the oxygen-to-metal ratio, the temperature, and the equilibrium oxygen partial pressure was proposed.

Enthalpy measurement on (UPu)O (x = 0, 0.18, 0.45, and 1) and analysis of heat capacity

Hirooka, Shun; Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*; Kato, Masato; Murakami, Tatsutoshi

Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09

https://doi.org/10.1016/j.jnucmat.2024.155188

no abstracts in English

Sintering behavior analysis of compacted dry recycled UPuO powder using master sintering curve theory

Nakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi

Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07

https://doi.org/10.1016/j.jnucmat.2024.155072

Estimating the corrosion rate of stainless steel R-SUS304ULC in nitric acid media under concentrating operation

Irisawa, Eriko; Kato, Chiaki

Journal of Nuclear Materials, 591, p.154914_1 - 154914_10, 2024/04

https://doi.org/10.1016/j.jnucmat.2024.154914

The amount of corrosion of austenitic stainless-steel R-SUS304ULC was evaluated considering the changes in solution composition and boiling during actual concentration operations. Austenitic stainless-steel R-SUS304ULC is the structural material of the highly radioactive liquid waste concentrator in Japanese spent fuel reprocessing plant, which treats highly corrosive nitric acid solutions during enrichment operations. The study results show that it is necessary to focus on nitric acid concentrations, oxidizing metal ion concentrations, and decompression boiling as factors that accelerate the corrosion rate of stainless steel because of cathodic reaction activation.

Dissolution behavior and aging of iron-uranium oxide

Tonna, Ryutaro*; Sasaki, Takayuki*; Okamoto, Yoshihiro; Kobayashi, Taishi*; Akiyama, Daisuke*; Kirishima, Akira*; Sato, Nobuaki*

Journal of Nuclear Materials, 589, p.154862_1 - 154862_10, 2024/02

https://doi.org/10.1016/j.jnucmat.2023.154862

The dissolution behavior of FeUO compounds formed by a high-temperature reaction of UO with iron, a stainless-steel component of reactor structural materials, was investigated under atmospheric conditions. The compounds were prepared in an electric furnace using UO and FeO as starting materials, and their solid states were analyzed using X-ray diffraction, scanning electron microscopy energy dispersive X-ray spectroscopy, and X-ray absorption fine structure spectroscopy. The concentration of nuclides dissolved in water was examined by performing static leaching tests of FeUO compounds for up to three months. A redox reaction was proposed to occur between trivalent Fe and pentavalent U ions in the early stage of FeUO dissolution. It was thermodynamically deduced that the reduced divalent Fe ion was finally oxidized into a trivalent ion in the presence of dissolved oxygen, and iron hydroxide limited the solubility of Fe. Meanwhile, the concentration of hexavalent U (i.e., uranyl ion) was limited owing to the presence of secondary minerals such as metaschoepite and sodium uranate and subsequently decreased, possibly owing to sorption on Fe oxides, for example. The concentrations of multivalent ions of fission products, such as Ru and Ce, also decreased, likely for the reason above. By contrast, the concentration of soluble Cs ions did not decrease. The validity of this interpretation was supported by comparing the results with the dissolution behavior of a reference sample (Fe-free UO).

Thermal conductivity measurement of uranium-plutonium mixed oxide doped with Nd/Sm as simulated fission products

Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01

https://doi.org/10.1016/j.jnucmat.2023.154799

The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with NdO/SmO, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT). The dependences of the coefficients A and B on the Nd/Sm content (C and C, respectively) are evaluated as: A(mK/W)=1.70 10 + 0.93C + 1.20C, B(m/W)=2.39 10.