Kousaku Fukuda

Fission product palladium-silicon carbide interaction in htgr fuel particles

Abstract Interaction of fission product palladium (Pd) with the silicon carbide (SiC) layer was observed in irradiated Triso-coated uranium dioxide particles for high temperature gas-cooled reactors (HTGR) with an optical microscope and electron probe microanalyzers. The SiC layers were attacked locally or the reaction product formed nodules at the attack site. Although the main element concerned with the reaction was palladium, rhodium and ruthenium were also detected at the corroded areas in some particles. Palladium was detected on both the hot and cold sides of the particles, but the corroded areas and the palladium accumulations were distributed particularly on the cold side of the particles. The observed Pd-SiC reaction depths were analyzed on the assumption that the release of palladium from the fuel kernel controls the whole Pd-SiC reaction.

Fission product release from ZrC-coated fuel particles during post-irradiation heating at 1800 and 2000°C

Abstract The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particles for high-temperature gas-cooled reactors. Post-irradiation heating tests of the ZrC-Triso coated UO 2 particles were performed at 1800°C for 3000 h and at 2000°C for 100 h to study the release behavior of fission products. The fission gas release monitoring and the X-ray microradiography revealed that no through-coating failure occurred during the heating tests. The high cesium retention of the ZrC-Triso coated fuel particles was confirmed up to 1800°C. The diffusion coefficient for cesium in the ZrC layer was more than two orders smaller than that in the SiC layer at 1800°C. The diffusion coefficient for ruthenium in the ZrC layer was almost the same as that for cesium in the SiC layer.

Chemical vapor deposition of silicon carbide for coated fuel particles

Abstract Silicon carbide was chemically vapor deposited on the pyrolytic carbon-coated fuel particles in the fluidized bed reactor using methyltrichlorosilane, hydrogen, and/or argon. The coating conditions were varied systematically and the deposits were examined by X-ray diffractometry. The deposits were found to be β-SiC, β-SiC + Si and β-SiC + C depending on deposition conditions. To understand the CVD processes, the thermodynamic analysis was made on the system of the experiment. The analysis showed that the deposit compositions at thermodynamic equilibrium were β-SiC and β-SiC + C under the experimental conditions. From these results, a model of the CVD processes was presented based on the mass transfer mechanism. This model explained the experimental results fairly well.

Fission product release from ZrC-coated fuel particles during postirradiation heating at 1600°C

Abstract Release behavior of fission products from ZrC-coated UO 2 particles was studied by a postirradiation heating test at 1600°C (1873 K) for 4500 h and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the test. Ceramographic observations showed no palladium attack and thermal degradation of ZrC. Fission products of 137 Cs 134 Cs, 106 Ru, 144 Ce, 154 Eu and 155 Eu were released from the coated particles through the coating layers during the postirradiation heating. Diffusion coefficients of 137 Cs and 106 Ru in the ZrC coating layer were evaluated from the release curves based on a diffusion model. 137 Cs retentiveness of the ZrC coating layer was much better than that of the SiC coating layer.

Release behavior of metallic fission products from HTGR fuel particles at 1600 to 1900°C

Abstract Release behavior of metallic fission products from the Triso-coated UO 2 particles was studied by postirradiation heating tests in the temperature range 1600 to 1900°C (1873 to 2173 K) and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the tests. Ceramographic observations showed no palladium attack and thermal decomposition of SiC, 137 Cs, 134 Cs, 110m Ag, 154 Eu and 155 Eu were released from the coated particles through the coating layers during postirradiation heating. The diffusion coefficient of 137 Cs in the SiC layer was evaluated from the release curves based on a simple diffusion model assuming a one-layer coated particle. Fractional release measurements suggested that the diffusion coefficient of 110m Ag in SiC be larger than that of 137 Cs.

Deterioration of ZrC-coated fuel particle caused by failure of pyrolytic carbon layer

Abstract The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particles for high-temperature gas-cooled reactors. To understand the behavior of the ZrC-Triso-coated fuel particles at 1800 to 2000°C, a ceramographic examination and an electron probe microanalysis were performed on the ZrC-Triso-coated fuel particles after the post-irradiation heating tests and a thermodynamic analysis of the ZrCUO system was carried out. Based on the results of the examination and analyses, a mechanism of the deterioration of the ZrC-Triso-coated fuel particles was proposed. The deterioration of the ZrC-Triso-coated fuel particles observed at 1800 to 2000°C was caused by the failure of the inner pyrolytic carbon layer.

Research and development of HTTR coated particle fuel

The coated particle fuel has been developed within a framework of the HTTR (High Temperature engineering Test Reactor) Development Program at the Japan Atomic Energy Research Institute. The HTTR fuel is a prismatic block type containing TRISO-coated U02 particles. Research and development on the fuel has been progressed in three categories; a work for fuel production technology, a proof test of fuel performance and a safety-related research. In the present report the concept and outline of the fuel in the HTTR design are firstly described, and then fuel fabrication technology including recently developed methods for improving fuel quality is followed. Tests for proving fuel performance have been carried out extensively on the reference fuel of the HTTR design by irradiation in an in-pile gas loop and capsules, and typical results are presented in this report. Concerning the safety-related research, fuel failure and 137Cs release at abnormally high temperature are described.

Fission product behavior in Triso-coated UO2 fuel particles

Abstract The behavior of fission products in irradiated Triso-coated UO 2 fuel particles was examined by electron probe microscopy, and the thermodynamic analysis was carried out on the fission products-UO 2 -C system to understand the fission product behavior in the coated particles. In the UO 2 kernels the precipitates of Mo, Pd-Te and Pd-Mo-Sn were observed besides the Mo-Tc-Ru-Rh-Pd alloy. In the coating layers palladium, tellurium, cerium and barium were often observed. The chemical form of barium and cerium was oxide, while the probable form of tellurium was elemental tellurium. The calculated vapor pressure of CeO 2 was the highest of the species containing the rare earth elements, and the calculated main Ba-containing gaseous species was BaO. The intercalation compound C n Cs was thermodynamically predicted to exist as a dominant chemical form of cesium at high temperatures instead of Cs 2 MoO 4 .

Structure of chemically vapour deposited silicon carbide for coated fuel particles

The silicon carbide coating layers prepared under various conditions were examined by density measurement, X-ray diffractometry, and optical and scanning electron microscopies in order to clarify the relation between deposition conditions and structure of the coating layers. It was found that the deposition temperature was the main parameter affecting the content of free silicon, density, crystallite size and lattice distortion, and microstructure. The dependence of these properties on the coating rate and the composition of fluidizing gas was not observed clearly. Free silicon was co-deposited withβ-SiC at temperatures lower than 1400 to 1500° C, and the content of free silicon increased with decreasing deposition temperature. The density of the layers without free silicon was more than 3.210 Mg m−3 and the density decreased with increasing content of free silicon. Crystallite size increased with deposition temperature and lattice distortion decreased with increasing deposition temperature. The outer surfaces of the layers without free silicon consisted of large interlocked grains, whereas those of the layers with free silicon showed a cauliflower-like structure of which the apparent grain size was small.

Solubility of magnesium in uranium dioxide

Abstract The solubility of magnesium in uranium dioxide under low oxygen pressures was studied at 1200°C. Magnesium was found to dissolve up to y > 0.1 (and below y = 0.15) of the apparent formula, MgyU1−yO2 + x ( x ⪋ 0 ) on heating at po2 = 10−15 and ≤ 10−19 atm. The formed solid solution in such a low po2 region was of the type (MgaU1−a){Mgb}O2 + c, in which the magnesium atoms partly occupy the interstitial sites together with the substitutional sites for uranium atoms. The ration of interstitial atoms to the total magnesium atoms increased from 0.23 (y = 0.05) or 0.39 (y = 0.1) at po2 = 10−15 atm with decreasing oxygen partial pressure to 0.62–0.63 (y = 0.05 and 0.1) at po2 ≤ 10−19 atm. The lattice parameter of the (MgaU1−a){Mgb}O2 + c solid solutions was represented as a linear equation of a, b and c. The interstitial magnesium caused an increase in the lattice parameter, in contrast to the substitutional magnesium which largely decreases the lattice parameter. It is possible that the uranium atoms in the solid solutions prepared at low oxygen partial pressures (≤ 10−19 atm) were reduced to slightly less than the tetravalent state.