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Papers by katsumi une

Journal of Nuclear Science and Technology, 1974

Journal of Nuclear Science and Technology, Aug 1, 1979

The relation between strain rate and iodine stress corrosion cracking (SCC) was studied on Zircal... more The relation between strain rate and iodine stress corrosion cracking (SCC) was studied on Zircaloy-2 subjected to uniaxial stress under constant extension rate in an iodine partial pressure of 4 Torr and at a temperature of 350°C. The specimens were machined from actual fuel cladding tube. In iodine environment, the tangentially directed specimens registered sharp decreases in stress beyond maximum point; this is attributed to crack initiation and propagation. Fracture ductility diminished with decreasing strain rate. Severe SCC was observed at strain rates below 2×10−3min−1. At high strain rates, the mechanism governing the rate of SCC appeared to be the time-dependent corrosion process. The axially directed specimens showed no signs of embrittlement due to gaseous iodine; this is attributed to the particular texture of the Zircaloy. Fracture surface observations indicated that transgranular cleavage fracture along the basal plane appeared to play a significant role in the stages of SCC initiation and p...

Journal of Nuclear Materials, Feb 1, 1991

Abstract Sintering kinetics of powder compacts of UO 2 -(5,10 wt%)Gd 2 O 3 and UO 2 have been stu... more Abstract Sintering kinetics of powder compacts of UO 2 -(5,10 wt%)Gd 2 O 3 and UO 2 have been studied in controlled atmospheres of H 2 O / H 2 and CO 2 / CO mixed gases by using a dilatometer. The densification rates and microstructure of the sintered pellets are considerably influenced by both the sintering atmosphere and Gd 2 O 3 content. After a heat treatment of 1650° C for 2 h, the sintered densities for UO 2 -Gd 2 O 3 pellets begin to decrease above the threshold oxidizing atmospheres, while the density for the UO 2 pellet increases slightly with more oxidizing atmospheres. These behaviors result from the difference in development of pore structure during sintering: the pore structure of UO 2 -Gd 2 O 3 pellets varies from an open pore structure to a closed pore structure on changing the sintering atmosphere from reducing to oxidizing. On the other hand, the pore structure of the UO 2 pellet is hardly affected by the sintering atmosphere. The formation of (U,Gd)O 2 solid solutions and the grain growth are enhanced with more oxidizing atmospheres.

Journal of Nuclear Materials, Oct 1, 1982

Partial molar thermodynamic quantities for urania-gadolinia solid solutions of compositions U1−yG... more Partial molar thermodynamic quantities for urania-gadolinia solid solutions of compositions U1−yGdyO2+x, with y values of 0.04 to 0.27, have been obtained using a solid electrolyte galvanic cell technique. The measurements were made for O/M ratios ranging from near stoichiometry to 2.20, and for temperatures ranging from 700 to 1100°C. The results for pure UO2+x are in accordance with data reported earlier. The oxygen potentials for U1−yGdyO2+x are higher than for pure UO2+x and increase positively with increasing Gd content or excess oxygen. They can be represented as a function of the mean U valence, except at the stoichiometric composition. Both the partial molar entropy and enthalpy increase negatively with increasing Gd content or excess oxygen.

Chemischer Informationsdienst, Jan 10, 1984

Progress in Nuclear Energy, 2012

Journal of Nuclear Materials, 2011

In situ diffusion experiments of the hydrogen isotope deuterium in the oxide layer formed on zirc... more In situ diffusion experiments of the hydrogen isotope deuterium in the oxide layer formed on zirconium alloys were carried out to clarify the hydrogen diffusion mechanism in the layer. The experiments were done in deuterium plasma for the temperature range from 523 to 673K by using a nuclear reaction analysis for D(3He,p)4He. The zirconium alloys used were GNF-Ziron (high iron

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), 1977

Journal of ASTM International, 2011

Journal of Nuclear Science and Technology, 2002

The effect of rim structure formation and external restraint pressure on fission gas release at t... more The effect of rim structure formation and external restraint pressure on fission gas release at transient conditions has been examined by using an out-of-pile high pressure heating technique for high burnup UO2 fuels (60, 74 and 90 GWd/t), which had been irradiated in test reactors. The latter two fuels bore a developed rim structure. The maximum heating temperature was 1500ºC, and the external pressures were independently controlled in the range of 10-150 MPa. The present high burnup fuel data were compared with those of previously studied BWR fuels of 37 and 54 GWd/t with almost no rim structure. The fission gas release and bubble swelling due to the growth of grain boundary bubbles and coarsened rim bubbles were effectively suppressed by the strong restraint pressure of 150 MPa for all the fuels; however the fission gas release remarkably increased for the two high burnup fuels with the developed rim structure, even at the strong restraint conditions. From the stepwise de-pressurization tests at an isothermal condition of 1500ºC, the critical external pressure, below which a large burst release due to the rapid growth and interlinkage of the bubbles abruptly begins, was increased from a 40-60 MPa level for the middle burnup fuels to a high level of 120-140 MPa for the rim-structured high burnup fuels. The high potential for transient fission gas release and bubble swelling in the rim-structured fuels was attributed to highly over-pressurized fission gases in the rim bubbles.

Journal of Nuclear Materials, 2009

Journal of Nuclear Materials, 2013

ABSTRACT In situ diffusion measurements of the hydrogen isotope deuterium in the oxide layers for... more ABSTRACT In situ diffusion measurements of the hydrogen isotope deuterium in the oxide layers formed on Zr–2.5Nb alloy have been carried out at 523 and 573 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis for the D(3He,p)4He reaction. The oxide layers were prepared in two environments, 1 M LiOH-containing water at 563 K and steam at 673 K, and their thicknesses ranged from 1.6 to 1.9 μm. The deuterium profile evolution in the oxides showed a combined process of absorption and dissolution in the surface region, and subsequent bulk diffusion in the deeper region. The diffusion coefficients of deuterium were evaluated for the two formation environments from the transient deuterium profiles of the oxide layers. The diffusion coefficients in the LiOH–water oxide were significantly larger than the values in the steam oxide. Compared with previously obtained data for three kinds of Zircaloy-type alloys (Zry-2, GNF-Ziron and VB), both of the Zr–2.5Nb oxides possessed the smallest diffusivities among the four alloys. Moreover, the surface concentrations of deuterium in the Zr–2.5Nb oxides were distinctly lower than those in the other Zircaloy-type alloy oxides. The superior hydrogen absorption performance of Zr–2.5Nb alloy observed in the out-of-pile corrosion tests was attributed to the smaller diffusivity and the smaller concentration gradient. The mechanism for the lower hydrogen diffusion flux of Zr–2.5Nb alloy was discussed in terms of the dissolution effect of niobium with variable valences of Nb2+–Nb5+ from the β-Nb precipitates into the ZrO2 lattice.