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Journal Articles

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

Times Cited Count：1 Percentile：72.91(Materials Science, Multidisciplinary)

The thermal conductivities of near-stoichiometric (U,Pu,Am)O $_{2}$ doped with Nd $_{2}$ O $_{3}$ /Sm $_{2}$ O $_{3}$ , 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) $^{-1}$ . The dependences of the coefficients A and B on the Nd/Sm content (C $_{Nd}$ and C $_{Sm}$ , respectively) are evaluated as: A(mK/W)=1.70 10 $^{-2}$ + 0.93C $_{Nd}$ + 1.20C $_{Sm}$ , B(m/W)=2.39 10 $^{-4}$ .

Journal Articles

A Raman spectroscopy study of bicarbonate effects on UO $_{2+x}$

McGrady, J.; Kumagai, Yuta; Watanabe, Masayuki; Kirishima, Akira*; Akiyama, Daisuke*; Kimuro, Shingo; Ishidera, Takamitsu

Journal of Nuclear Science and Technology, 60(12), p.1586 - 1594, 2023/12

https://doi.org/10.1080/00223131.2023.2236104

Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)

Journal Articles

Oxygen potential, oxygen diffusion, and defect equilibria in UO $_{2 pm x}$

Watanabe, Masashi; Kato, Masato

Frontiers in Nuclear Engineering (Internet), 1, p.1082324_1 - 1082324_9, 2023/01

https://doi.org/10.3389/fnuen.2022.1082324

Since the oxygen potential and the oxygen coefficient of UO $_{2}$ have a significant impact on fuel performance, many experimental data have been obtained. However, experimental data of the oxygen potential and the oxygen diffusion coefficient in the high temperature region above 1673 K are very limited. In the present study, we aimed to obtain these data and analyze them by defect chemistry. The oxygen potentials and the oxygen chemical diffusion coefficient of UO $_{2}$ were measured by the gas equilibrium method in the near stoichiometric region at temperatures ranging from 1673 to 1873 K. A data set of oxygen potentials was made together with literature data and analyzed by defect chemistry. The oxygen potential of UO $_{2}$ was determined as a function of O/U ratio and temperature, and an equation representing the relationship was derived. The oxygen chemical diffusion coefficient values obtained in this study were reasonably close to the literature values. The oxygen partial pressure dependence of the oxygen chemical diffusion coefficients was predicted from the evaluated results of the oxygen potential data, but no clear dependence was observed.

JAEA Reports

Evaluation of the minimum critical amount for heterogeneous lattice systems composed of fuel rods utilized in low-power water-moderated research and test reactors by using continuous-energy Monte Carlo code MVP with JENDL-4.0

Yanagisawa, Hiroshi

JAEA-Technology 2021-023, 190 Pages, 2021/11

JAEA-Technology-2021-023.pdf:5.25MB

Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.

Journal Articles

Chlorination of UO $_{2}$ and (U,Zr)O $_{2}$ solid solution using MoCl $_{5}$

Sato, Takumi; Shibata, Hiroki; Hayashi, Hirokazu; Takano, Masahide; Kurata, Masaki

Journal of Nuclear Science and Technology, 52(10), p.1253 - 1258, 2015/10

https://doi.org/10.1080/00223131.2015.1070110

Times Cited Count：6 Percentile：45.92(Nuclear Science & Technology)

In order to explore the applicability of the chlorination by MoCl $_{5}$ as a potential pretreatment technique for waste treatment of fuel debris by pyrochemical methods, chlorination experiments of UO $_{2}$ and (U $_{0.5}$ Zr $_{0.5}$ )O $_{2}$ simulated fuel debris were carried out in two steps: the first one is a chlorination reaction by homogeneous heating, the second one is a volatilization of molybdenum by-product by heating under temperature gradient condition. Most of UO $_{2}$ and (U $_{0.5}$ Zr $_{0.5}$ )O $_{2}$ powder were converted to UCl $_{4}$ or UCl $_{4}$ and ZrCl $_{4}$ mixture at 573 K, respectively. In the case of (U $_{0.5}$ Zr $_{0.5}$ )O $_{2}$ sintered particle, most of sample was converted to the chlorides because the products evaporated and be separated from sample surface at 773 K, while only the surface of the sample disk was converted to the chlorides at 573 and 673 K. Most of molybdenum by-product and ZrCl $_{4}$ were separated from UCl $_{4}$ by volatilization at 573 K.

JAEA Reports

Essentials of neutron multiplicity counting mathematics; An Example of U-Pu mixed dioxide

Hosoma, Takashi

JAEA-Research 2015-009, 162 Pages, 2015/08

JAEA-Research-2015-009.pdf:22.3MB

Neutron coincidence counting assay systems have been developed in the last two decades. Objects would extend to high-mass uranium-plutonium dioxide containing other spontaneous fission nuclei, so essentials of neutron multiplicity counting were reconsidered and expanded: (a) Formulae of multiplicity distribution were algebraically derived up to septuplet using a probability generating function; (b) Leakage multiplication was evaluated not by Monte Carlo method but by an average length from an arbitrary point inside a sample to an arbitrary point on its surface and a probability of induced fission within the length; (c) Mechanism of coincidence counting was associated with a couple of different time axes in Poisson process, and consequently a pair of close-to-coincident neutrons from the process was derived. For the formulae, new expressions using combination were wrote down. For spectrum and mean free path, actually treated uranium-plutonium dioxide was selected as an example.

JAEA Reports

Report on the fuel treatment facility operation

Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu

JAERI-Tech 2005-004, 53 Pages, 2005/03

JAERI-Tech-2005-004.pdf:5.92MB

This report presents operating records of dissolution of uranium dioxide and concentration of uranyl nitrate solution and acid removal, which have been performed from 1994 through 2003, for the purpose of feeding 10% and 6% enriched uranyl nitrate solution fuel to Static Experimental Critical Facility(STACY) and Transient Experimental Critical Facility(TRACY) in Nuclear Fuel Safety Engineering Facility(NUCEF).

Journal Articles

An Investigation into dissolution rate of spent nuclear fuel in aqueous reprocessing

Mineo, Hideaki; Isogai, Hikaru; Morita, Yasuji; Uchiyama, Gunzo*

Journal of Nuclear Science and Technology, 41(2), p.126 - 134, 2004/02

https://doi.org/10.3327/jnst.41.126

Times Cited Count：7 Percentile：45.11(Nuclear Science & Technology)

A simple equation was proposed for the dissolution rate of spent LWR fuel, of which the change in the dissolution area was estimated by taking into account of the area of the cracks occurring due to thermal shrinkage of the pellets during irradiation. The applicability of proposed equation was examined using LWR fuel dissolution test results in the present study as well as the results obtained by other workers. The equation showed good agreements with the dissolution test results obtained from spent fuel pellets and pulverized spent fuel. It was indicated that the proposed equation was simple and would be useful for the prediction of dissolution of spent LWR fuels. However, the initial effective dissolution area, the parameter of the equation, was found to depend on the temperature, which could not be explained by the proposed equation. Further studies on the role of other factors affecting dissolution rate, such as nitrous acid, in the dissolution of spent fuel was required.

Journal Articles

Investigation of vaporization behavior of hyper-stoichiometric uranium dioxide by Knudsen effusion mass spectrometry

Nakajima, Kunihisa; Arai, Yasuo

Journal of Nuclear Materials, 317(2-3), p.243 - 251, 2003/05

https://doi.org/10.1016/S0022-3115(03)00111-9

Times Cited Count：3 Percentile：25.79(Materials Science, Multidisciplinary)

The Knudsen effusion mass-spectrometric measurement of pure UO $_{2+x}$ (s) are carried out at 1673, 1773 and 1873K to evaluate G $_{f}$ $^{circ}$ (UO $_{3}$ ,g) as well as to measure the partial pressures of UO $_{3}$ (g) and O $_{2}$ (g) over UO $_{2+x}$ (s) as function of the O/U ratio. It was found that the partial pressures of O $_{2}$ (g) over UO $_{2+x}$ (s) almost agree with the experimental data reported in the past and the values derived from the empirical equation given by Nakamura and Fujino. Further, it was found that the values of G $_{f}$ $^{circ}$ (UO $_{3}$ ,g) obtained in this study are in good agreement with the recommended values.

Journal Articles

Behavior of simulated spent fuel in subcritical water

Mineo, Hideaki; Suzuki, Tadashi; Morita, Yasuji

Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.334 - 338, 2003/00

Behavior of spent nuclear fuel in subcritical water was investigated to look at the feasibility of fission-products (FPs) separation without organic solvent. The study employed unirradiated UO $_{2}$ particles simulating spent fuel burned up to 45,000MWdt $^{-1}$ , which includes FP elements in oxide form: Sr, Zr, Mo, Ru, Rh, Pd, Ag, Ba, La, Ce, Pr, Nd and Sm. Also, alloy particles consisted of Mo, Ru, Rh and Pd were prepared to simulate the metallic phase of FP. 12.728 g of the fuel and 52 mg of the alloy were placed in a 10 ml pressure vessel, where subcritical water was fed. The temperature was 523, 573, 623 and 663K, while the pressure was kept at 29MPa. Dissolved fraction decreased with elevating temperature. It was found that more than 5% of Ba, Mo and Pr were respectively dissolved. The dissolved fraction of Sr and Rh were about 1%, and about 0.3% for Zr. La, Ce, Nd and Sm, indicated almost the same result as U, which was about 0.1%. It was suggested that the subcritical water could separate portion of FP. Further study would be carried out with smaller-sized fuel.

JAEA Reports