Sachio Shimada

A Simple Model for Hydrogen Re-distribution in Zirconium-lined Fuel Claddings

A simple model of hydrogen diffusion and hydride precipitation was developed to estimate the radial re-distribution of hydrogen in zirconium-lined Zircaloy-2 claddings for light water reactors. Numerical calculation based on the model agreed well with experimental data and clearly showed that a slight difference of the terminal solid solubility of hydrogen in the two materials caused a significant flow of hydrogen from Zircaloy-2 to zirconium during slow cooling. The model would also evaluate the hydrogen distribution in the claddings of which the terminal solid solubility is not uniform.

Evaluation of the resistance of irradiated zirconium-liner cladding to iodine-induced stress corrosion cracking

Abstract An evaluation was made of irradiated zirconium-liner cladding for its resistance to iodine-induced stress corrosion cracking (SCC). Emphasis was put on irradiation-induced hardening in zirconium and SCC resistance in zirconium-liner cladding as compared with Zircaloy-2 cladding. The Vickers microhardness test revealed that crystal bar zirconium experienced less hardening than Zircaloy-2 during neutron exposure. The SCC resistance of zirconium-liner cladding was evaluated for failure strains under the tube pressurization SCC test, and compared with the results of Zircaloy-2 cladding. The failure strains of zirconium-liner cladding were significantly larger than those of Zircaloy-2 cladding over all neutron fluence ranges examined, e.g., more than ten times at 1.0 × 10 21 n / cm 2 ( E > 1 MeV ). Judging from our results on the Vickers microhardness and SCC tests, good SCC resistance of zirconium-liner cladding could be expected even at high fluences.

Development of new zirconium alloys for a BWR

Specimens for irradiation testing in a commercial BWR were prepared from 24 kinds of Zr alloys. The corrosion performance of these specimens was investigated after irradiation for up to four cycles. Two cycles of irradiation were not always sufficient to identify the lowest corrosion alloys. Marked acceleration of corrosion occurred for many alloys between one and four cycles. The onset of mildly accelerated oxidation was observed for standard Zircaloy-2 at four cycles of irradiation. On the other hand, for Zircaloy-like alloys, increasing Fe and Ni contents or decreasing the Sn content promoted a saturation oxidation between two and four cycles of irradiation. The effects of alloying elements on corrosion behavior were evaluated by both in-pile and out-of-pile corrosion tests. The alloying elements Fe, Cr, and Ni, which have smaller valences than Zr, improved the in-pile corrosion resistance of Zr alloys, while the alloying elements Nb, Mo, and Te, which have larger valences than Zr, were responsible for increased weight gain during the irradiation test in the BWR.

Stress Corrosion Cracking of Zircaloy-2 by Metal Iodides

Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method. The results showed that: (1) only those metal iodides which react thermodynamically with Zr to produce ZrI4 cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility; (3) gaseous ZrI4 is the most corrosive agent; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I2 vapor. Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI4, formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.

Effect of Hydrogen Water Chemistry on Zircaloy-2 Fuel Cladding and Structure Material Performance

The effects of hydrogen addition to the feedwater on the corrosion and hydrogen uptake performance of Zircaloy-2 fuel cladding tubes, a water rod tube and spacer materials irradiated for four cycles in a BWR were evaluated. The uniform oxide behaviors of the cladding tubes, water rod and spacer materials were not affected by hydrogen water chemistry (HWC) condition. The hydrogen uptake and pickup fractions of the water rod and spacer materials were similar to those of water rods and spacer materials under normal water chemistry (NWC) conditions. As for the fuel rods, in spite of comparably heavy crud deposition, their hydrogen uptake and pickup fractions were clearly lower than the values under NWC conditions. Overall, the results indicated that HWC had no adverse effects on fuel performance.

Experimental Study on Fluoride Volatility Process for Thorium Fuels

Selective removal of uranium from (Th/U)O2 by fluorination with fluorine was studied experimentally. The fluorination was performed both in a small boat and in a 2 inch inner diameter fluid-bed reactor. Fuel particles tend to agglomerate in the reactor due to the large amount of reaction heat and the comparatively low melting point of ThF4. The fluorinated fuels produced in the fluid-bed reactor were found to be partially agglomerated. Fractional retention of uranium was smaller in the agglomerated parts than in the un-agglomerated, and smaller in the outer layers of the cakes than in the core. On the other hand, it was also established beyond doubt from the results of the small boat fluorination experiment that heavy agglomeration inhibits the volatilization of uranium in the form of UF6. Inhibition of the violet exothermic reaction by lowering the fluorine pressure in the early stage of fluorination was found to be a very effective method of obtaining high uranium recovery. It was demonstrated that more...