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A study was made of uranium contamination in (a) the coating layers of TRISO particles (a-1) before compacting and (a-2) separated from once-compacted fuel heat-treated at 1,400 or 1,800°C, and (b) in the matrix material of the same compacts. The contamination in the pyrocarbon layers of the coating was determined, after mechanically separating the coating layers, by a procedure of neutron activation, burn-off and 133Xe trapping. For the silicon carbide coating layer, the fragments of coating left from the above procedure were fused into alkaline melt, and the 133Xe released at each heating step was trapped. For the matrix material, the fuel compacts were deconsolidated electrolytically or mechanically, followed by activation analysis. The results of the foregoing measurements proved the uranium contamination in pyrocarbon and silicon carbide coating to be at most of the order of 10−4 in reference to uranium content in kernel, while the corresponding value for particles sampled from fuel compacts heat-tre...

Axial, circumferential and radial distributions of fission products in the graphite sleeve, inner-tube and block of irradiated high temperature gas-cooled reactor (HTGR) fuel assemblies were measured by gamma spectrometry, lathe sectioning and beta counting with ion-exchange separation. Some distinctive peaks of 144Ce and 125Sb in their axial profiles, together with the very high activity level of fission products are ascribed to the failure of coated fuel particles. The effective retention capability of the graphite sleeve was observed for 90Sr, 106Ru, 125Sb, 144Ce and 155Eu; whereas not for 134Cs and 137Cs. Silver-110m was detected in graphite materials of the fourth OGL-1 fuel assembly with an increased burnup of 1.96% fissions per initial metal atom (FIMA). Effective in-pile diffusion coefficients of 90Sr, 125Sb and 144Ce in the graphite sleeves have been estimated using the Fickian diffusion theory.

The fission product release from TRISO-coated UO 2 particles was measured by post-activation heating at 1940 to 2320°C for use in a safety analysis. The results are analyzed mathematically with effective diffusion coefficients in each medium. 103 Ru, 99 Mo and 95 Nb are released at 1940 to 2320°C and have high effective diffusion coefficients. Although 140 Ba and 137 Cs are retained in TRISO-coated particles at 2050°C, they are released rapidly at 2320°C. This is attributed to the transition of beta to alpha SiC at 2320°C. 141 Ce, 140 La and 95 Zr are released little if any at 2320°C. Rare gas nuclides, iodine and tellurium seem to be retained in coated particles at this high temperature.

Distribution of metallic fission products in the graphite sleeve and block of the fifth OGL-1 fuel assembly was measured by gamma spectrometry with lathe sectioning. Considerably large release fractions of long-lived fission products with smooth axial profiles were observed in the sleeve due to a large failure fraction of coated fuel particles accompanied with failed silicon carbide layers. Nevertheless, a key nuclide110mAg, whose large release is suspected at increased burnups for low-enriched uranium fuels, was effectively retained within the graphite sleeve. The retention was also observed for125Sb, 154Eu and155Eu up to a burnup of 3.2% fission per initial metal atom, but was limited for134Cs and137Cs at high sleeve-temperatures above 900°C. In-pile diffusion coefficients in IG-110 graphite have been evaluated for Cs, Ag and Sb; those for Cs are in reasonable agreement with available in-pile data.

Diffusion coefficients of Ce, Sr and Ba in the matrix graphite for VHTR fuel compact have been determined by measuring the radial profiles of 141Ce, 89Sr and 140Ba concentrations in slightly irradiated graphite spheres which have U02 kernels at the center. Temperature dependences of the diffusion coefficients of Sr and Ba have been determined between 1,175 and 1,375°C, and they are expressed as where T is the absolute temperature and R the gas constant (8.314 J/mol·K). The diffusion coefficient of Ce in the above temperature range was rather higher than those of Sr or Ba, and the magnitudes of these coefficients were expressed in following order; DCe>DSr>DBa. The effect of matrix graphite density on diffusion of the fission products was studied, but difference was not found clearly in the diffusion coefficients of Ba which had been obtained in both of low and high density matrices at 1,375°C.

Triso-coated UO 2 particles with ZrC as the third layer were heated at 2173–2773 K. The particles withstood the heating at 2723 K for 1 h, though more than one half failed at 2773 K within 1 h. The results were compared with those on the conventional Triso-coated particles with SiC as the third layer. The thermodynamics of the system within ZrC-Triso particles are discussed and the internal CO pressure at high temperatures is estimated. The results of the impurity analysis on CVD ZrC is also presented.

Simultaneous deposition of carbon and zirconium from vapor produced by the reaction between methyl iodide vapor and zirconium sponge was studied with the application of a spouted bed constituted by a funnel carrying a charge of alumina microspheres, which were blown upward and held in dynamic suspension by a jet of the vapor and gases spouting from the funnel. The purpose of the experiment was to determine the conditions favorable for obtaining a coat of zirconium carbide-carbon alloy on the microspheres. Deposition of the vapor on the microspheres, leading to the formation of the carbon alloy coating, was found to take place at temperatures exceeding 1,100°C. The C/Zr ratio of the deposited coat was found to increase with deposition temperature. The hydrogen concentration in the spouting gas affected both the deposition yield and the chemical composition of the deposit. Repeated use of the sponge was found to impair its performance due to deactivation by premature deposition of carbon.

The whole-particle and third-layer crushing strengths of SiC-Triso and ZrC-Triso coated fuel particles were measured. The third-layer crushing strengths did not differ for the two types of Triso-coated particles if their particle diameters and third-layer thicknesses were not different. The third-layer failure stresses calculated with the model by Delle et al. [1] were found still dependent on the particle radius. The failure stress divided by the particle radius was about constant for a given third-layer material. The whole-particle crushing strength of SiC-Triso particle was larger than that of ZrC-Triso particle in the as-coated condition, but the contrary was the case when those particles were annealed at 2073 K for 1 h. By the annealing the whole-particle strength of SiC-Triso particle decreased but that of ZrC-Triso particle increased. Both of these changes could not be ascribed to the corresponding changes in the third-layer strength.

ZrC coatings by chemical vapor deposition were applied to particles of ThO 2 , UO 2 and Al 2 O 3 at 1623–1873 K. The feed gas mixture consisted of ZrBr 4 , CH 4 , H 2 and Ar. The results were compared with the calculated chemical equilibria in the Zr-C-H-Br system. It was shown that the weight and composition of the deposit can be calculated by thermochemical analysis after correcting the methane flow rate for a pyrolysis efficiency. Predominant reaction presumably occurring were derived by a mass balance consideration on the calculated equilibrium species. A simplified model of the ZrC deposition was proposed.

In order to develop the fuel of a high temperature gas-cooled reactor, axial and radial profiles of fission and activation products in the graphite sleeves of the first and second fuel assemblies irradiated in the in-pile gas loop OGL-1 were measured by means of lathe sectioning, gamma spectrometry, and ion-exchange separation. Several sharp axial profile peaks of 90Sr, 134Cs, and 137Cs were observed for the second fuel sleeves. The peaks are assigned to the failed coated fuel particles; the rest of the profiles to the contamination uranium contained in the fuel compact matrix. Fission products with lower diffusivity such as 90Sr, 106Ru, and 144Ce were completely retained in the second fuel sleeve, whereas similar effectiveness in retention was not observed for 134Cs and 137Cs. Diffusion coefficients of cesium and strontium have been roughly estimated through the comparison between the measured and the simply calculated profiles in the sleeve.

Laboratory scale experiments have been carried out to obtain information on the sources of CH3I, reported by various workers to have been detected under accidental conditions of fuels. Small pieces of slightly irradiated sintered UO2 and U metal turnings were heated externally up to 1,400° C in various carrier gases, and the carrier gases were analyzed for fission product iodine present in the form of CH3I. From the data obtained, it is concluded as follows: In the case of sintered UO2, some amount of CH3I appears to be released directly from the fuel below 600°C, whereas in case of U metal turnings, direct release does not apparently occur at any temperature. In oxidizing atmospheres, iodine released from either sintered UO2 or U metal is considered to change into CH3I only to a small extent, probably by gas phase formation.

微細ポアー構造の研究には従来もっぱら窒素吸着法が行なわれてきたが, 液体窒素温度でクリプトン吸着を行なうと, 従来法より低圧で行ないうるために精度を上げることができる. この考え方は比表面積測定に生かされている. これを微細ポアー構造の研究手段にまで拡張することを実験的に検討した. 例として黒鉛の酸化およびUO2の焼結に伴なう微細ポアー構造の変化を調べた.

In connection with a program to study the behavior of punctured fuel elements for the Tokai Atomic Power Reactor, the diffusion coefficient of fission gas in uranium oxide powder formed by CO2 oxidation of U was determined by post-irradiation experiment, in which the fractional release of fission gas during isothermal heating of the powder was measured. The U was oxidized at 600° and 700°C, and in both cases the O/U ratio of the oxides, measured gravimetrically, was 2.0. The diffussion coefficients in the oxide powder formed by oxidation at 600°C were found to be 1.4× 10-20, 1.3×10-19, 1.1×1018 and 1.0×10-17, cm2sec-1, respectively at 450°, 550°, 650° and 750°C, and in the oxide powder formed at 700°C, 7.4×10-19 and 3.6×10-16cm2sec-1 at 600° and 700°C, respectively. Activation energies calculated for the two oxide powders were comparatively low.

Release of Xe from lightly sintered UO2 was measured by post-irradiation heating in vacuum at 400°~800°C. The release rate after initial burst followed the diffusion model. An Arrhenius plot of the diffusion coefficients revealed a discontinuity at 600°C. Activation energies were calculated to be 3.4 kcal/mol at temperatures lower than the discontinuity, and over 36kcal/mol in higher temperature range, which corresponded respectively to non-lattice and lattice diffusion. The discontinuity is explained by a changeover of the predominance of one process over another depending on temperature range. The initial burst coald not be attributed to surface oxidation. Based on analyses of such factors as dependence of burst on temperature and on specific surface area, as well as percentage of burst release, a kind of activated extrication from superficial layer, approximately 5A thick, was proposed as a concept for interpreting the burst phenomenon.