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Yeasts generally grow with a highly glycolytic metabolism and restrain mitochondrial biogenesis except for some Fe/S proteins. Respiratory mitochondrial functions and biosynthesis pathways are well studied, however how cells coordinate basal fermentative mitochondrial functions is not fully understood. We show that a part of nuclear-encoded mitochondrial mRNAs, which preferentially translated in monosomes, are regulated by Puf3p upon glucose-rich media. Especially, of those monosome-enriched nuclear-encoded mitochondrial mRNAs, CAT5/COQ7 mRNA has a variant of the canonical Puf3p binding site on its 3’-UTR. Western blot analysis showed that Puf3p represses the translation of Cat5p regardless of fermentable or respiratory media. In vitro binding assay revealed that Puf3p directly binds to CAT5 mRNA via the non-canonical Puf3p binding site.Mutants harboring the substitution of the non-canonical Puf3p binding site in CAT5 mRNA grew normally but impaired Cat5p expressions apparently, indicating CAT5 mRNA is a bona fide Puf3p target. Overall, Puf3p, a general key modulator for nuclear-encoded mitochondrial mRNAs, fine-tunes translation of a subset of nuclear-encoded mitochondrial mRNAs including mRNAs with non-canonical Puf3p binding sites under the fermentation. This may be required to keeping the fundamental functions of yeast mitochondria at proper levels.

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.

The concentration distributions of 133 Xe, 140 Ba, 89 Sr, 141 Ce, 103 Ru and 95 Zr- 95 Nb in a SiC layer and releases of these fission products from SiC coated fuel particles were measured in a temperature range from 1650° to 1850°C, to obtain the diffusion parameters and to investigate the diffusion behaviors in SiC. Temperature dependences of the diffusion coefficients of these fission products except 95 Zr- 95 Nb were obtained. In diffusion behaviors of the alkali earth fission product, volume diffusion occured considerably besides grain boundary above 1650°C. 141 Ce and 103 Ru diffusions were almost through grain boundaries in the temperature range. 95 Zr- 95 Nb showed volume diffusion besides boundaries diffusion in their distribution annealed at 1850°C for 35 h, but those for the shorter time or at the lower temperature would be due to boundary diffusion. In comparison of the diffusion coefficients in SiC and PyC, it was proved that SiC was effective for retention of the solid fission products.

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.

Mixtures of ZrC-C were chemically vapour deposited from gaseous mixtures of zirconium bromides, methane, hydrogen and argon. The effect of gas composition on the deposition behaviour was studied. The experiments have shown that the methane concentration in the feed gas mixture is a crucial factor in determining the deposition rate and the character of the deposit. Chemical equilibria in the Zr-C-H-Br system were calculated and compared with the experimental results.

The diffusion behavior of 137 Cs, 90 Sr and 144 Ce in CVD-ZrC was studied by measuring concentration profiles in ZrC and inner PyC layers of the coated particles, and the release from the particles during post-irradiation annealing. 90 Sr diffusion in ZrC is found to be controlled by a volume diffusion mechanism at 1600°C; this is found also for 137 Cs although it is influenced by micro-pores. The diffusion coefficient of 137 Cs obtained from the analysis of the concentration profiles in ZrC 1.3 is D = 2.3 × 10 −3 exp( −3.18 × 10 5 RT ) ( cm 2 s ) , where D is the diffusion coefficient, R the gas constant (8.314 J/mol · K) and T the temperature (K). In case of 144 Ce, the volume diffusion is not so remarkable as 90 Sr.

The distributions of 137 Cs, 90 Sr and 144 Ce in the coating layers of several kinds of TRISO coated fuel particles for HTGR were determined by method of removing the coating layers stepwise after irradiation. The distributions in SiC coating layer are influenced by its density and irradiation temperature. High dense SiC coating layer showed good retention of the fission products at such high irradiation temperature as 1530°C, while the retention of low dense one would change at 1300°C; above the temperature, it would be lost. There are several patterns of the distribution in outer PyC coating layer such as high accumulation of the fission products in a certain position of the layer, outerward rises in the concentrations and high concentrations in the layer than those in some positions of SiC coating layer. The post irradiation annealing was also carried out for one kind of the particle.

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.

メタンの空気による部分酸化によって生ずる残留メタン-水蒸気-水素-酸化炭素-炭酸ガス-窒素の各成分比を,温度900,1000および1200°Kの条件で,空気-メタン混合比を変数として化学平衡論の立場から計算した。またこの反応には炭素析出の可能性があるが,この境界条件も温度および空気-メタン混合比を変数として算出した。以上の結果から全反応領域を4種に区分して考察をおこなった。計算上おこなった仮定の成否を試験するため,天然ガスと空気の混合ガスをニッケル触媒を用い,(750℃,a(O2/CH4)=0.68),(800℃,a=0.68),(950℃,a=0.80)等の条件で反応させ,生成したガスの各成分百分率をもとめ計算値との比較をおこなった。その結果によれば両者の間に矛盾がみとめられなかったので,計算の前提および得られた計算結果は正しいものと考えられる。また炭素析出領域である750℃,a=0.13における反応を実験的にしらべ,非炭素析出領域における結果との比較をおこなった。

The escape of 133Xe introduced into natural graphite by recoil from UO2 particle was studied at temperatures ranging from 200° to 1,450°C and the resulting data correlated with the duration of isothermal post-irradiation heating. The escape data were used to verify the proposed escape mechanism, which postulates three successive processes—release from trapping site, diffusion of the released xenon, and desorption from surface. The concept of initial activation energy spectrum, which determines the trapping states of xenon atoms before any annealing begins in terms of the activation energy for the release, was introduced to express the release behavior mathematically. The mathematical treatment for both the release and the diffusion processes permits expression of the whole escape process with a single equation, assuming no resistance to xenon transport in the desorption process. The validity of this equation was substantiated by observed data. It was made clear that rapid sweep of the spectrum is the caus...

Purex 1 WW型模擬廃液に正リン酸を加えて, 1100℃に加熱してP2O5含量が65重量%のリン酸ガラスを造り, これを用いて, 蒸留水, 水道水および海水による137CsとPとの浸出を調べた。その結果, (1) 137Csの浸出は蒸留水によるよりも水道水や海水によるほうが多かったが, Pの浸出は水の種類にあまり影響されない, (2) 浸出量は温度とともに急激に増大する, (3) 浸出測定結果は, ガラス成分と水との反応における活性化エネルギーが広い範囲に分布する場合の解析を適用することにより, 活性化エネルギー初期スペクトルによって整理できる, などが明らかにされた。

An attempt has been made to pursue the course of the thermal decomposition of uranyl nitrate hexahydrate in the presence of graphite, and compositions of the product at various stages in the decomposition have been determined by X-ray diffraction and by thermogravimetric measurement. The influence of graphite on the decomposition appeared in the temperature region above 500DC, and reduction with graphite of the decomposed product into uranium dioxide was observed above 750°C. It was also demonstrated that uranium dioxide and carbide-dispersed graphite-base fuel could be prepared by thermal decomposition of hexahydrate in the graphite matrix.

In the past few years, great interest has been aroused in the HTGR from the viewpoint of effective use of nuclear energy as well as alleviation of environmental problems in Japan. This article outlines the coated particle fuels for the HTGR, and reviews the current reseach status of the relation between the properties and irradiation behavior of the coated fuel particles.

Using tracer technique, studies were undertaken on the adsorption of iodine on natural graphite at temperatures ranging from 27° to 900°C under pressures from 0.03 to 1.4 mmHg, and on the desorption of adsorbed iodine from the same specimens into either vacuum or Ar stream. The specimens were heated up either isothermally, stepwise, or at constant rate of temperature rise, up to 1,100°C. The resulting data accorded well with a theoretical treatment presuming desorption from heterogeneous surface, considering the desorption as an n-th order irreversible reaction having distributed activation energies. The iodine is considered to be held on the graphite by physical adsorption below about 450° C, and above that temperature, by chemisorption through bond formation between iodine and carbon atoms. The heat of adsorption and the activation energy of desorption obtained was found to be a function of surface coverage. The presence of Ar gas was observed to inhibit chemisorption of iodine and to decrease the activ...

Coated fuel parrticles with free Si in the SiC layer and UC in the UC/ sub 2/ kernel were heated to 1,180 to 1,700 deg C for up to 200 h. Examination was made by metallography, x-ray diffraction, and electron probe. Results show that pores are formed in the coating upon heating, that the content of Si and UC decreased with temperature, and that Si diffused into the inner PyC layer and U diffused toward the coating. The coated fuel particles are liable to deterioration by heating above 1,460 deg C. The possible reaction of U with SiC is suggested. (DLC)

The fission gas escape during fissioning in fuel was interpreted with a newly proposed escape mechanism which comprises the three successive processes of release from trapping sites, diffusion of the released fission gas nuclides, and desorption from the fuel surface. The expression for the release rate of fission gas was derived from a simple mathematical model postulating a continuous activation energy spectrum for the release reaction of trapped nuclides. The proposed mechanism satisfactorily explains observed data reported in literature on the dependence of the escape rate on decay constant, the contribution of the precursor, and the change of escape rate with irradiation time. There appears to be only a small effect brought by diffusion of the released nuclides on the escape rate, which can thus be considered to be controlled mainly by the release reaction from the the trapping sites. The present model based on the release from trapping sites can also qualitatively explain the influence of differences in irradiation conditions such as time and neutron flux.

再処理廃液の蒸発濃縮に液中燃焼法を適用する場合を想定して, プロパンと酸素の混合ガスを燃料とした液中燃焼法によりおもにPurex1WW型模擬廃液を蒸発濃縮する実験を行なった。その結果, (1) 通常の蒸発法に比べて缶液温度は10～16℃低く, しかも高度の濃縮が可能で減容比0.3まで濃縮できる, (2) 蒸発缶は小型で簡単な構造を有し熱効率は大きい, (3) 燃焼ガスからCO2の吸収を行ない, かつ, バーナーの不完全燃焼を少なくすれば排出ガス量はかなり少なくできる, (4) 硝酸アルミニウムを多量に含む廃液に対しては通常の蒸発によるよりも高度の濃縮が可能で蒸発終了後冷却すればただちに固型物が得られる, などが明らかにされた。